NMP1L3010, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

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Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
ML15089A233
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 03/26/2015
From: Jim Barstow
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15089A234 List:
References
NMP1L3010, TSTF-425, Rev. 3
Download: ML15089A233 (68)
v • d • e


Text

TABLE 4.6.2a INSTRUMENTATION THAT INITIATES SCRAM

@1..t l lw,'.a Dnu,4. m iramari4 aM LAIEI I W

L-IMI I5 V

IU&7L II Parameter (1)

Manual Scram (2)

High Reactor Pressure (3)

High Drywell Pressure (4)

Low Reactor Water Level (5)

High Water Level Scram Discharge Volume (6)

Main-Steam-Line Isolation Valve Position (7)

Deleted None None None Noneeday None None Instrument I

Channel Test Onee PeF We O"..

pe. 3 M 1) nee peFi fUelthSMl Onoc peF 3 mointhSO Instrument I Channel Calibration None Onee peF 3 FMM8)

Onee peF3 menthe~1 )

Gnee peF 3 mfentI990)

Onec peF -3 mognths Gnoc per 3 monthe A

I Unee pcr eseroting ovoic I

AMENDMENT NO. 44, 149 201

TABLE 4.6.2a (cont'd)

INSTRUMENTATION THAT INITIATES SCRAM Paramete (8)

Shutdown Position of Reactor Mode Switch (9)

Neutron Flux (a) IRM (I)

Upscale (11) Inoperative (b) APRM (i)

Upscale (11) Inoperative (10) Turbine Stop Valve Closure (11)

Generator Load Rejection Surveillance K(Note1 7

Sensor Check None G-ee per shift(r oNeeoPe None Instrument Channel Test

)nce during each major refueling outage Instrument I

Channel Calibration None O.ee.pe week(g)

.Ree.Pe week(g)

Onoc por 3 months Onoc per 3 IAM1nth Onoc por 3 monAth8 Ono3 pcr 3 mon~tho On.o po. opo;ating

,yee (n)

Onco por oporating eyl. n Gnee-pef week 1

. n e e.p eF.. f ;.,..,. (n)

None None None None Onee aer onoratinA evelec Ono-p, *3 m*enth AMENDMENT NO. 4421463-186 202

NOTES FOR TABLES 3.6.2a and 4.6.2a (a)

May be bypassed when necessary for containment inerting.

(b)

May be bypassed In the refuel and shutdown positions of the reactor mode switch with a keylock switch.

(c)

May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi, or for the purpose of performing reactor coolant system pressure testing and/or control rod scram time testing with the reactor mode switch in the refuel position.

(d)

No more than one of the four IRM Inputs to each trip system shall be bypassed.

(e)

No more than two C or D level LPRM inputs to an APRM shall be bypassed and only four LPRM inputs to an APRM shall be bypassed In order for the APRM to be considered operable. No more than one of the four APRM inputs to each trip system shall be bypassed provided that the APRM in the other instrument channel in the same core quadrant is not bypassed. A Traversing In-Core Probe (TIP) chamber may be used as a substitute APRM input if the TIP is positioned in close proximity to the failed LPRM it is replacing.

(f)

Verify SRM/IRM channels overlap during startup after the mode switch has been placed in startup. Verify IRM/APRM channels overlap at least 1/2 decade during entry into startup from run (normal shutdown) if not performed within the previous 7 days.

(g)

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before startup, if not performed within the previous 7 days. Not required to be performed during shutdown until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering startup from run.

(h)

Each of the four Isolation valves has two limit switches. Each limit switch provides input to one of two instrument channels in a single trip system.

(I)

May be bypassed when reactor power level is below 45%.

(j)

Trip upon loss of oil pressure to the acceleration relay.

(k)

May be bypassed when placing the reactor mode switch in the SHUTDOWN position and all control rods are fully inserted.

(1)

OnlY the trip 81rOUlt will be oal ibratod and tooted at the frequonoicsopepifiod in Table 4.6.2a, the prir cocrWell bo calibratod and tosted onec per operating eyolo.

ý S Rý (m)

This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during reactor operation when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust the APRM channel if the difference is greater than +2.0/-1.9% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 2.1.2a shall not be included in determining the difference.

(n)

Neutron detectors are excluded.

AMENDMENT NO. 4427 44,4T4" 4 186 203

NOTES FOR TABLES 3.6.2a and 4.6.2a (n)

Deleted.

(o)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same trip system is monitoring that parameter.

With one channel required by Table 3.6.2a inoperable in one or more Parameters, place the inoperable channel and/or that trip system In the tripped condition* within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With two or more channels required by Table 3.6.2a inoperable in one or more Parameters:

1.

Within one hour, verify sufficient channels remain Operable or tripped* to maintain trip capability for the Parameter, and

2.

Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, place the inoperable channel(s) in one trip system and/or that trip system** in the tripped condition*, and

3.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, restore the inoperable channels in the other trip system to an Operable status or tripped*.

Otherwise, take the ACTION required by Specification 3.6.2a for that Parameter.

An Inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, If the Inoperable channel is not restored to Operable status within the required time, the ACTION required by Specification 3.6.2a for the parameter shall be taken.

This ACTION applies to that trip system with the most inoperable channels; if both trip systems have the same number of inoperable channels, the ACTION can be applied to either trip system.

(p)

May be bypassed during reactor coolant system pressure testing and/or control rod scram time testing.

AMENDMENT NO. 142 204

TABLE 4.6.2b INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Requirement N~UM I )

Sensor Checkl Instrument Channel Test Instrument Channel Calibration PRIMARY COOLANT ISOLATION (Main Steam, Cleanup and Shutdown Cooling)

(1)

Low-Low Reactor Water Level vGei/dey Gne 8F3fnntqd)

Onoc during eaeh major Fefueliftg eut GRee P8 3 menhe(d)

(2)

Manual MAIN-STEAM-LINE ISOLATION (3)

High Steam Flow Main-Steam Line Oieelday Gnee peF3 M1499(d)

Onee peF 3 F~en(Sd)

(4)

Deleted (5)

Low Reactor Pressure GVeiVday Gnee peF-3 M19 d)

Onee peF 8 FflOAqthqd)

AMENDMENT NO. 4-42, 44, 197 209

w TABLE 4.6.2b (cont'd)

INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Reaulrement 7

Instrument Channel Calibration Permmea Selnsr Check Charnrnel T (6)

Low-Low-Low Condenser Vacuum (7)

High Temperature Main-Steam-Line Tunnel CLEANUP SYSTEM (8)

High Area Temperature SHUTDOWN COOLING SYSTEM ISOLATION None None unee aunng coon mejor FeW*~

es.A*49ll.s*.

tW,. ~v,.l,.

nna ""*n Anah vma,,ar Onoc during each m~ajor Onoc duNR@n eaeh mjor Fef,.

"Rg I4 _.,*

v Fefueling eut"e I

Gneelweek Mama dwrina angah nut-6w Fefueling ewtW I

Gnoc during eaoh major Fef% lffi I4,,; _^ *

(9)

High Area Temperature G.eewAeek unec Dar ooemtn~ o';oie unec per opcrating eyele I I

AMENDMENT NO. 442, 177

TABLE 4.6.2b (cont'd)

INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Reaulrement eo CInstrument SesrCe Channel Test Parameter CONTAINMENT ISOLATION (10)

Low-Low Reactor Water Level (11)

High Drywall Pressure (12)

Manual Instrument I

Channel 2

Calibration

.. ee pe, 3 me h (d) peF3 f,.le.,.S (d)

Onee/day Gneelday Gnee peF-8menthq d Gnee PeF 3 fnehSd)

AMENDMENT NO. 142 211

NOTES FOR TABLES 3.6.2b and 4.6.2b (a)

May be bypassed In the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi.

(b)

May be bypassed when necessary for containment inerting.

(c)

May be bypassed in the shutdown mode whenever the reactor coolant system temperature is less than 2151F.

(d) el tc ri aralt will be oal-brated and tested at the froqucnoico spcoifiod in Table 4.6.2b, the primnar; soncor Will be oulibroo tootod on por prting eyeolo.

(e)

Delete INSERT3 (f)

A channel may be placed In an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system, either

1.

Place the inoperable channel(s) in the tripped condition within

a.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Parameters common to SCRAM Instrumentation, and

b.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Parameters not common to SCRAM Instrumentation.

or

2.

Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for both trip systems,

1.

Place the inoperable channel(s) in one trip system in the tripped condition within one hour.

and

2.
a.

Place the Inoperable channel(s) in the remaining trip system in the tripped condition within (1) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Parameters common to SCRAM Instrumentation, and (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Parameters not common to SCRAM Instrumentation.

or

b.

take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142 212

NOTES FOR TABLES 3.6.2b and 4.6.2b (g)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels for the Operable Trip System, either

1.

Place the Inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or

2.

Take the ACTION required by Specification 3.6.2a for that Parameter.

(h)

Only applicable during startup mode while operating in IRM range 10.

(i)

May be bypassed in the cold shutdown condition.

(J)

In the cold shutdown and refueling conditions, only one Operable Trip System is required provided shutdown cooling system integrity is maintained. With one of the two required Operable Channels in the required Trip System not operable, place the inoperable channel in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Otherwise, either:

1. Immediately Initiate action to restore the channel to operable status.

or

2. Immediately Initiate action to isolate the shutdown cooling system.

F Note 1: SurveIIanoe Intervals are specified In the Surelac uFrequency Control Program unles otheriwse noted in thal AMENDMENT NO. 442,4 3,197 213

TABLE 4.6.2c INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING Surveillance Reauirement ensorICh Ie Sensor Check Parmete Instrument Channel Test Instrument Channel Calibration EMERGENCY COOLING INITIATION (1)

High Reactor Pressure (2)

Low-Low Reactor Water Level EMERGENCY COOLING (for each of two systems)

(3)

High Steam Flow Emergency Cooling System None 0nela None AMENDMENT NO. 142 215

NOTES FOR TABLES 3.6.2c AND 4.6.2c (a)

Each of two differential pressure switches provide inputs to one instrument channel in each trip system.

(b)

May be basdinte cold shutdown condition.

(c) ~I~~~fibe eaflbrated and tested at the frequcneic3 speoified in Table 4.6.2e, the primarfy Gceore will be oalibrated and totedonce er oerting eyele.

(d)

A channel may be placed in an Inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

(e)

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1.

For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the action required by Specification 3.6.2a for that Parameter.

2.

With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

(f)

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system, either

1.

Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or

2.

Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for both trip systems,

1.

Place the inoperable channel(s) in one trip system in the tripped condition within one hour and

2.
a.

Place the inoperable channel(s) in the remaining trip system in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or

b.

Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142 216

TABLE 4.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Surveillance Reaulrement pt~wim 1/

SensorCheck Pmaraer Instrument Channel Test Instrument Channel Calibration START CORE SPRAY PUMPS (1)

High Drywell Pressure (2)

Low-Low Reactor Water Level OPEN CORE SPRAY DISCHARGE VALVES (3)

Reactor Pressure and either (1) or (2) above Qnee1day Gneeklay Gnee peF8 menhsc)

Gnee peF 8Mea h "()

Onee peF, fleh

~C)

Onee peF3 F~nsc)

None Gnee p8F3 rmntsc)

GRee peF,3 Hi~t99c)

AMENDMENT NO. 142 218

NOTES FOR TABLES 3.6.2d AND 4.6.2d (a)

(b)

(c)

(d)

(0)

(f)

May be bypassed when necessary for containment inerting.

May be bypassed when necessary for performing major maintenance as specified in Specification 2.1.1.e.

1ni t

'4i --!t will be oallbmted and tested at the froquene-es speofficd in Table 4.6.2d, the prirnor;y scnor w.ill be ealibratod 01d-t461 n

  • n~

~

r'AFAAA May be bypassed when necessary for integrated leak rate testing.

The Instrumentation that Initiates the Core Spray System is not required to be operable, if there is no fuel in the reactor vessel.

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1.

With one channel Inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.

2.

With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142 219

TABLE 4.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Surveillance Parameter (1)

a.

High Drywell Pressure Sensor Check 16 onee/day Instrument Channel Test O.ee peF 3 FMe4h(b)

Intww Ien Instrument Channel Calibration

- neigths(b)

b.

Low-Low Reactor Water Level Gieeky*a Gie pF3 eth b)

Gnee peF 3 ffeth b)

AMENDMENT NO. 142 221

NOTES FOR TABLES 3.6.29 AND 4.6.2e (a)

Mae assed In the shutdown mode whenever the reactor coolant temperature is less than 215 0F.

(b)

.. I tl Vp oreult will be ealibr-ated and tcotcd at the ffequenoies spcoificd In Table 4.6.2e, the pimr IFoo Will be oalIbrotcd an t-to onc pe operating OYGoeo.

(c)

A channel may be placed in an Inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip system in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1.

With one channel Inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.

2.

With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142 222

TABLE 4.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION Surveillance Reaulrement Sensor CheckV Instrument 7 Channel Te!st Instrument Channel Calibration INITIATION (1)

a.

Low-Low-Low Reactor Water None GR~e eF,3 M119(c)

GRee PeF3 M8119(0 and

b.

High Drywell Pressure GIeel*ay Onee peF 8 c)

Gnee peF 8 149(c)

AMENDMENT NO. 142 224

NOTES FOR TABLES 3.6.2f AND 4.6.2f (a)

B Instrument channels In eithe trip system are required to be energized to initiate auto depressurization. One trip system is powered from power board 102 and the other trip system from power board 103.

(b)

May be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding

saturation temperature.

E (C)

Unl thrFpol~ult wll~ be oallbratod and tested at the froquenoioa spoolfied in Table 4.6.2f, the prmArFY SBRcor Will be calibrntod and tested onoc par oporating eyole-.

(d)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.
2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142 225

TABLE 4.6.2g INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Surveillance Reouirement Instrumelnt Sensr CeckChannel Tes Parameter (1)

Deleted (2)

IRM

a.

Detector not in Startup Position

b. Inoperative
c.

Downscale

d.

Upscale (3)

APRM

a.

Inoperative

b.

Upscale (Biased by Recirculation Flow)

c.

Downscale (4)

Deleted N/A N/A N/A N/A None None None

"-^,...,, pe: weekt(g) on-"e pe: week(g)

Gnee p" wee 9(g)

"-^^ peF week(g)

Onoo por 3 FAMonth Gnoc por 3 mone~th Onco por 3 Monfth Instrumne~nt Channel Calibration N/A N/A

.n..

pe. eperting eyeleJ)

Onc pcFr pr*ating eyeleJ)

None Oe-. pm-3 menhe0 )

Onee peF 3 mentfhe)

AMENDMENT NO. 186 230

TABLE 4.6.2g (cont'd)

INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Surveillance Reaulrement Instrum Channel Test Instrumenti Channel Calibration~i p-4um

1) 171, Sensor Check Parmete (5)

Refuel Platform and Hoists (6)

Mode Switch In Shutdown (7)

Mode Switch in Refuel (Blocks withdrawal of more than 1 rod)

(see 4.5.2)

Onee during eeehmao Onee during eechmao r-euein eu.:.

(8)

Deleted AMENDMENT NO. 442186 231

NOTES FOR TABLES 3.6.2g and 4.6.2g (a)

Deleted (b)

No more than one of the four IRM inputs to each instrument channel shall be bypassed. These signals may be bypassed when the APRMs are onscale.

(c)

No more than one of the four APRM inputs to each instrument channel shall be bypassed provided that the APRM in the other Instrument channel in the same core quadrant is not bypassed. No more than two C or D level LPRM inputs to an APRM shall be bypassed and only four LPRM inputs to only one APRM shall be bypassed in order for the APRM to be considered operable. In the Run mode of operation, bypass of two chambers from one radial core location in any one APRM shall cause that APRM to be considered Inoperative. A Travelling In-Core Probe (TIP) chamber may be used as a substitute APRM input if the TIP is positioned in close proximity to the failed LPRM It Is replacing. If one APRM in a quadrant is bypassed and meets all requirements for operability with the exception of the requirement of at least one operable chamber at each radial location, it may be returned to service and the other APRM In that quadrant may be removed from service for test and/or calibration only if no control rod is withdrawn during the calibration and/or test.

(d)

Deleted (e)

Deleted (f)

One sensor provides input to each of two instrument channels. Each instrument channel is in a separate trip system.

(g)

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before startup, if not performed within the previous 7 days. Not required to be performed during shutdown until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering startup from run.

(h)

The actuation of either or both trip systems will result in a rod block.

(I)

A channel may be placed In an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the Trip System in the tripped condition, provided at least one other operable channel in the same Trip System is monitoring that Parameter.

(J)

Neutron detectors are excluded.

x Note 1: Surveillace Intervals aer speciffed In the Surveillance Frequency Control Program unless otherwise~ notedi.the AMENDMENT NO. 442466, 211 232

TABLE 4.6.2h VACUUM PUMP ISOLATION Surveillance Yp 'w w I I)

Sensor ~hoký Instrument Channel Test Instrument Channel Calibration Parameter MECHANICAL VACUUM PUMP High Radiation Main Steam Line lnee lqa Onec pc, 3

.O.thS Onoc por 3 months AMENDMENT NO. 142 235

NOTES FOR TABLES 3.6.2h and 4.6.2h (a)

Deleted.

(b)

A channel may be placed In an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system, either

1.

Place the Inoperable channel(s) in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

or

2.

Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for both trip systems,

1.

Place the Inoperable channel(s) in one trip system in the tripped condition within one hour.

and

2. a. Place the Inoperable channel(s) in the remaining trip system in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

or

b. Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142 236

TABLE 4.6.21 DIESEL GENERATOR INITIATION Surveillance hDNe1)'~7 IIww 1)f Instrument(`)

Channel Test Instrument(')

Channel Calibration Parmete Loss of Power

a.

4.16kV PB 102/103 Emergency Bus Undervoltage (Loss of Voltage)

b.

4.16kV PB 102/103 Emergency Bus Undervoltage (Degraded Voltage)

NA Onc8 por rofuoling cyelo G-ov por FeofUling eyeln NA (a)

The Instrument channel test demonstrate the operability of the instrument channel by simulating an undervoltage condition to verify that the tripping logic functions properly.

(b)

The Instrument channel calibration will demonstrate the operability of the instrument channel by simulating an undervoltage condition to verify that the tripping logic functions properly. In addition, a sensor calibration will be performed to verify the set points listed in Table 3.6.2.1.

(c)

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

~Note 1: Surveilance Intervast arespecifid In the SureUincFrequenyentrqrgram unles tewientdi h AMENDMENT NO. 142 239

TABLE 4.6.2j EMERGENCY VENTILATION INITIATION IkIUL0 I )

Sensor C!hc Surveillance Reauirement I(oEI1)

Instrument Channel Test '

Channel Calibration Paacmete (1)

High Radiation Reactor Building Ventilation Duct (2)

High Radiation Refueling Platform GRee/ehift (b)

On,.

during e8Gh oporating y.,,

(C)

AMENDMENT NO. 142 241

NOTES FOR TABLES 3.6.21 AND 4.6.21 (a)

This function shall be operable whenever recently irradiated fuel or an irradiated fuel cask is being handled in the reactor building, and dtions with a potential for draining the reactor vessel (OPDRVs).

(b )

" " ---"- th is f.._ _,1_,_ req u ired to b e ep r_,.l1 (c)

Immediately prior to when function is required and enee pep we" thereafter until function is no longer required.

(d)

A channel may be placed In an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels for the Operable Trip System, either

1) Place the Inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or

2)

Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 442, 194 242

TABLE 4.6.2k HIGH PRESSURE COOLANT INJECTION Parameter (1)

Low Reactor Water Level (2)

Automatic Turbine Trip G(Noeepme7 Surveillance Requirement Instrument

  • Channel Test Pd r

3 FA onti, ib)

  • nee d'-*n-. eah ep.rati... e'ele Instrumentj Channel Calibration O n eepeF (b)

None None

"*W wj w.

AMENDMENT NO. 142 244

NOTES FOR TABLES 3.6.2k AND 4.6.2k (a)

May be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation ter aure.

(b)~~

~~~

ldl ic1,Ofl ilb alibratod and testod at tho froguoniele spoolficd in Table 4.6.2k, the

'OFar cecrWill be eal-bratod and tootod onoc pcr opcratlng cyelc.

(c)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.
2.

With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142 245

TABLE 4.6.21 CONTROL ROOM AIR TREATMENT SYSTEM INITIATION Surveillance Sensor Check Instrument Channel Test Instrument /

Channel j Calibration Onee pe euft q

b)

(1)

Low-Low Reactor Water Level (2)

High Steam Flow Main-Steam Line (3)

High Temperature Main-Steam Line Tunnel Gneeishlf Gneeiday Gnee pef quatb)

Gnee pef qta~eb)

Gee pe q elb)

Gnee eaoh operating oyole not to emocod 24 mon~ths Gnot to.h cpcrating oyol.

not to oxoood 24 menthe (4)

High Drywell Pressure GGneeday Gnee pe que(tb)

GneePef qukeb)

AMENDMENT NO. 442, 161 247

NOTES FOR TABLES 3.6.21 AND 4.6.21 (a)

May be bypassed when necessary for containment inerting.

NSERT3 (b)

o.

.rut will bc ealibrated and tested at the ffeguenoi il be cairated and tested onee (c)

May be bypassed in the cold shutdown condition.

AMENDMENT NO. 464, 194 247a

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION High Flow-Main Steam Line, +1 psld High Flow-Emergency Cooling Line, +/-1 psid High Area Temperature-Main Steam Line, +/-10 0F High Area Temperature-Clean-up and Shutdown, +/-60F High Radiation-Main Steam Line, +100% and -50% of set point value High Radlatlon-Reactor Building Vent, +100% and -50% of set point High Radiation-Refueling Platform, +100% and -50% of set point Spoolield eur,'lllanoc Intcrmais and awurciianee and maintonenoo outago fimces haveo boen doterminod in acoorda nee with N EDC 30861 PA "Toohnical Spcciflcation Imgproyomcnt Analyeea for BWR Rcaotor P~etcctien Systemn," MDE T? 0485, "Teehnical Speoifioation fimprovamoint Anal~

wf Nine Mile Point Nuolear Stion, Unit 1," and Gcnori

,ottc, 91 04, a

iThniea Speoifieation Intomla to Aoaommodato a 21 Month Gyeoe.

Spoolfied oeoellllaneII intealal and eul*lIilanel and maintonanoo outage timo haIl beon doteominod in aocVidanee with NEDC 30861 P A Supl "Toohinlcai Spocifloation Improvornont Analyses for BWR isolation lnetwumontation Common to RPS and EGGS Instfumentation," and with NEDC 31677P A, "TIohnie-l SpcoifiIatiAn Improvement Analyses for BW.AIR l"olation Atuation inst"dmontation." Bocauso of loal high radiation, tooting inotrumentatlon in the aroa of the main otoam line isolation valvoc oan only be done durng poriods of Station chutdown. These funotions Inoludo high Mec tomqperaturo loolatlen and isolation volvo pocitien comm.

Ys 6 Fuel pi 2, AMENDMET*-

MN.-,

142,176, Revision 13 252

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION Spcolflcd eur~clllanoo intoryalo and survoilleneo and maintenanoc outage times have been determined in aooordanec with NEMDC 30936P A,A "BWR Ownoro' Group Toonoi pofoainiprvmn Mtooog.. h oocmto o B..RECS.ouanIotucttln, ore1ad n iLUf Tofno e

eiiatien Improvemont Menalyolog (wit De flc LmorgenoyF W urmGG AetongR System Actuatlion InOtrUMOntation for Nina Mile Point Nuoloar Station, Unit V.

Spoolfiod eurwoillanoo Intoryalo and eur~oIlianoo and maintonaneo outage times have boon deteFrmined in aooordonco with GENE 770 S ~oflotion,"

o a~d bthz N.RC and doournented in the SER (leftetert R. D. Binz IVfrom. E. Roes' dated July 21, 1902).

Testing of the scram uu ith the shutdown position of the mode switch can be done only during periods of Station shutdown since it always involves a scram.

b.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR is maintained greater than the SLCPR. The trip logics for these functions are 1 out of n; e.g., any trip on one of the eight APRM's, eight IRM's or four SRM's will result in a rod block. The minimum instrument channel requirements provide sufficient instrumentation to assure the single failure criteria Is me Slpe iid ouryoillanoo intorvalo and ouvoillanop and moaintonaneo outago timoc hove boon determfined in aeoordanreo GErE 0ý"

061 ao for Changoc to Surveillance Toot Intorwale and Allowed Out Of Soryieo Times for Seleeted Inotrumentation E-RTS 3

'1 ?'i-"htlenr," as apprei'i by the PIR-and' deemete ih rn1 the S

('rfte~ tr R. D' BWn-I" ffem G. E. Rossi danted u

..., *ei_.*._*.v,.

, u

'"ir "

Le.eF 9

.,,, "Ghanges in Teehnieal p

i en... lU te,

a 24..

AIL L.

The APRM rod block trip Is flow biased and prevents a significant reduction in MCPR especially during operation at reduced flow.

The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the SLCPR.

The APRM rod block also provides local protection of the core; i.e., the prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst case single control rod withdrawal error has been analyzed and the results show that with the specified trip settings rod withdrawal is blocked before the MCPR reaches the SLCPR, thus allowing adequate margin. Below -60% power the worst case withdrawal of a single control rod results in a MCPR > SLCPR without rod block action, thus below this level it is not required.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the SLCPR.

A.A.A.

G-NT NO'-. 14.2, Revision 13 253

r-MSTA'"] bli, dral I 97fal-71ral -1 A -7.1b iral ý I

-Q"PVr-11 I Akirl: PPQ"1PPU1=NT JkLM,.U I~

I EI

~

I I

U I--

-S 3.6.3 EMERGENCY POWER SOURCES 4.6.3 EMERGENCY POWER SOURCES Applies to the operational status of the emergency power sources.

Oblective:

To assure the capability of the emergency power sources to provide the power required for emergency equipment in the event of a loss-of-coolant accident.

a.

For all reactor operating conditions except cold shutdown, there shall normally be available two 115 kv external lines, two diesel generator power systems and two battery systems, except as further specified in "b," "c," "d," "e," and "h" below.

Applicability:

Applies to the periodic testing requirements for the emergency power sources.

Obiective:

To assure the operability of the emergency power sources to provide emergency power required in the event of a loss-of-coolant accident.

Specification:

The emergency power systems surveillance will be performed as indicated below. In addition, components on which maintenance has been performed will be tested.

a.

During each maior refueling outage - test for automatic startup and pickup of load required for a

]--

loss-of-coolant accident.

b.

K

- manual start and operation at rated load shall be performed for a minimum time of one hour.

Determine the specific gravity of each cell.

Determine the battery voltage.

b.

One 115 kv external line may be de-energized provided two diesel-generator power systems are operable. If a 115 kv external line is de-energized, that line shall be returned to service within 7 days.

AMENDMENT NO. 142 255

I WITIh~n r-nl~nITIflM FQR QPPRATIQN SURVEILLANCE REQUIREMENT

-fl-S555 S

  • ~h io 7Lc I
c.

One diesel-generator power system may be Inoperable provided two 115 kv external lines are energized. If a diesel-generator power system becomes Inoperable, it shall be returned to an operable condition within 14 days. In addition, if a diesel-generator power system becomes Inoperable coincident with a 115 kv line de-energized, that diesel-generator power system shall be returned to an operable condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d.

If a reserve power transformer becomes inoperable, It shall be returned to service within seven days.

a.

For all reactor operating conditions except startup and cold shutdown, the following limiting conditions shall be In effect:

(1)

One operable diesel-generator power system and one energized 115 kv external line shall be available. If this condition is not met, normal orderly shutdown will be Initiated within one hour and the reactor will be In the cold shutdown condition within ten hours.

'FI

c.
  • determine the cell voltage and specific gravity of the pilot cells of each battery.
d.

Surveillance for startup with an inoperable diesel-generator - prior to startup the operable diesel-generator shall be tested for automatic startup and pickup of the load required for a loss-of-coolant accident.

e.

Surveillance for operation with an inoperable diesel-aenerator - If a diesel-generator becomes inoperable from any cause other than an inoperable support system or preplanned maintenance or testing, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, either determine that the cause of the diesel-generator being inoperable does not impact the operability of the operable diesel-generator or demonstrate operability by testing the operable diesel-generator. Operability by testing will be demonstrated by achieving steady state voltage and frequency.

AMENDMENT NO. 442, 447, 179 256

BASES FOR 3.6.3 AND 4.6.3 EMERGENCY POWER SOURCES If a battery system becomes Inoperable, that battery system must be returned to an operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the 24-hour allowed outage time cannot be met, then Specification 3.0.1 must be entered immediately. The second paragraph of Specification 3.0.1 provides two options:

1. Place the unit in a condition consistent with the individual specification; however, in this case, the individual specification (i.e., 3.6.3) does not provide any action to take when a battery system has been inoperable for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. To determine required actions and action completion times, the Individual specifications for the systems supported by the battery system should be entered and reviewed to determine applicable actions. If no actions are applicable for the given reactor operating condition, then no actions are required.

or

2. Place the unit in an operational condition in which the specification is not applicable (i.e., cold shutdown).

Revision 28 258a

I IUITIklrb 1%t*1K1n1T1 rd Pno PFRATInN I

IMT~hfl

~fl~flI~fl PARADPATItJSURVEILLANCE REOUIREMENTS 3.6.5 Radioactive Material Sources Applies to the limit on source leakage for sealed or start-up sources.

Oblective:

To specify the requirements necessary to limit contam-ination from radioactive source materials.

Specification:

1. The leakage test shall be capable of detecting the presence of 0.005 microcurle of radioactive material on the test sample. If the test reveals the presence of 0.005 microcurie or more of removable contam-Ination, It shall Immediately be withdrawn from use, decontaminated and repaired or be disposed of in accordance with Commission regulations. Sealed sources are exempt from such leak tests when the source contains 100 microcurles or less of beta and/

or gamma emitting material or 10 microcuries or less of alpha emitting material.

2. Results of required leak tests performed on sources, if the tests reveal the presence of 0.005 microcurie or more of removable contamination, shall be reported within 90 days.

4.6.5 Radioactive Material Sources Applicabilitv:

Applies to the periodic testing requirements for source leakage.

Obiective:

To assure the capability of each source material container to limit leakage within allowable limits.

Specification:

Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the Commission or an agreement State, as follows:

1.

Each sealed source, except start-up sources subject to core flux, containing radioactive material, other than hydrogen 3, with a half-life greater than 30 days and in any form other than gas shall be tested for leakage and/or contaminatio at int...a. met to oxoood ri* months.

nao ldqFlT I

AMENDMENT NO. 142 265

I WITIKI 1%MK1n1T1nK1 OnD MDOPATInki I

IUTI~f t~f~flI~flM~AD DPDAIAMR"iRVFII I AN(rE REOUIREMENTS

3.

A complete Inventory of radioactive by-product materials, exceeding the limits set forth in 10 CFR 30.71, In sealed sources In possession shall be maintained current at all times.

2.

The periodic leak test required does not apply to sealed sources that are stored and not being used.

The sources excepted from this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.

3.

Start-up sources shall be leak tested wihi 31.days prior to being subjected to core flux and following any repair or maintenance.

AMENDMENT NO. 142 266

BASES FOR 3.6.5 AND 4.6.5 RADIOACTIVE MATERIAL SOURCES The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the probable leakage from the source material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. Quantities of Interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Parts 30.11-20 and 70.19. Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.

AMENDMENT NO. 142 267

I WITIM 1"nkInITIMN OnD nPOPATIntJ

-Q"RVPII I AMrF RPn"IRPUPNT I IUTIM~ ffl~fIT~lhI~AD lD~AT~lhi~IID~III A(~F FAIID~~II 3.6.11 ACCIDENT MONITORING INSTRUMENTATION Applies to the operability of the plant instrumentation that performs an accident monitoring function.

Obiective:

To assure high reliability of the accident monitoring instrumentation.

a.

During the power operating condition, the accident monitoring instrumentation channels shown in Table 3.6.11-1 shall be operable except as specified in Table 3.6.11-2.

4.6.11 ACCIDENT MONITORING INSTRUMENTATION Applicability:

Applies to the surveillance of the instrumentation that performs an accident monitoring function.

Obiective:

To verify the operability of accident monitoring instrumentation.

Specification:

Instrument channels shall be tested and calibrate*-et-I=--

I-APý

-L

,UL I*_

Z; i 08rrous u oa

!;P

n4

=Rw rni N.;;

AMENDMENT NO. 142 268

TABLE 4.6.11 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Channel Test £6 Instrument Channel Calibration Instrument Channel Calibration (1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

Deleted Deleted Reactor vessel water level Drywell Pressure Monitor Suppression Chamber Water Level Monitor Deleted Containment High Range Radiation Monitor Suppression Chamber Water Temperature Onee peF quake Gnee peF FR Gnee peF quake Onee peF M Gnee peF FIA Afn-p A1rinn, rn-h FA';We Fettierlmr euitsee Onee during eseh mnajer refueling outage Onez during oach ma~jer refueling euag Onee during each m~ajer refueling eutaeg AMENDMENT NO. 442, 494, 205 272

BASES 3.6.11 AND 4.6.11 ACCIDENT MONITORING INSTRUMENTATION Accident monitoring Instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," NUREG-0737, "Clarification of TMI Action Plan Requirements,"

November 1980, NUREG-0661, "Safety Evaluation Report Mark I Containment Long Term Program," and the NRC Final Rule, "Combustible Gas Control In Containment," made effective October 16, 2003 (68 FR 54123).

6poofifod euryellianoc intorwaio and euryoi~anoc and maintonanoc outago tim'"

haye boon dotcRminod in eooordanoo with GENE 770 "Bases 1fo Changee to 8urvoilllano Teat intewaie and Aloewod Out Of Servioc Times for Solcotcd lnctrurncntation Teohnioai Speeifleationo,"

ac approved by the NRC and documontod in tho SER (lottor to R. D. Biniz IV from C. E. Rocci datod July 21, 10)W.

Action 3a and Action 4a of Table 3.6.11-2 require that with the number of OPERABLE channels less than the total Number of Channels shown in Table 3.6.11-1, a Special Report must be prepared and submitted to the NRC within 14 days following the event. The term "event" refers to the reason that an instrument channel is inoperable. For the purpose of applying Action 3a and Action 4a of Table 3.6.11-2, removal of a single accident monitoring instrumentation channel from service for the sole purpose of performing routine TS required surveillances Is not considered an event requiring preparation and submittal of a 14-day Special Report. If a single accident monitoring Instrumentation channel is removed from service for other activities (e.g., to perform preventive maintenance), or if a channel fails, these events require preparation and submittal of a Special Report in accordance with Actions 3a and 4a.

AMF..'DM_'T,NO. *2,. Revision 44, 46,A491). 19 273

LIMITIN InnNnITInN FnR PFR TIM LIMII~fl~Ot~lITflN OR APRATOMSURVEILLANCEW REOUIREMENT 4

3.6.12 REACTOR PROTECTION SYSTEM AND REACTOR 4.6.12 REACTOR PROTECTION SYSTEM AND REACTOR TRIP SYSTEM POWER SUPPLY MONITORING Applies to the operability of instrumentation that provides protection of the reactor protection system and reactor trip system.

Oblectlve:

To assure the operability of the Instrumentation monitoring the power to the reactor protection system and reactor trip system.

Soeclfication:

a.

Except as specified In specifications b and c below, two protective relay systems shall be operable for each power supply.

TRIP SYSTEM POWER SUPPLY MONITORING Applicability:

Applies to the surveillance of instrumentation that provides protection of the reactor protection system and reactor trip system.

Obiective:

To verify the operability of protection instrumentation monitoring the power to the reactor protection and reactor trip buses.

Specification:

INSERT I a.

Demonstrate operability of the overvoltage, undervoltage and underfrequency protective instrumentation by performing an instrument channel test. This instrument channel test will consist of simulating abnormal power conditions by applying from a test source, an overvoltage signal, an undervoltage signal and an under-frequency signal to verify that the tripping logic up to but not including the output contactors functions properly.

AMENDMENT NO. 142 274

LIMITING CONDITION FOR OPERATION

] SURVEILLANCE REQUIREMENT

\\ý.

I -

A.

b.

With one protective relaying system inoperable, restore the Inoperable system to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove the power supply from service.

c.

With both protective relaying systems inoperable, restore at least one to an operable status within 30 minutes or remove the power supply from service.

b.

-Z-1:

I-Demonstrate operability of the overvoltage, undervoltage and underfrequency protective instrumentation by performing an instrument channel test. This instrument channel test will consist of simulating abnormal power conditions by applying from a test source an overvoltage signal, an undervoltage signal and an underfrequency signal to verify that the tripping logic including the output contactors functions properly at least once. In addition, a sensor calibration will be performed to verify the following setpoints.

ii.

iii.

Overvoltage Undervoltage Underfrequency

_5132 volts, _<4 seconds

_>108 volts, *<4 seconds

> 57 hertz, <2 seconds AMENDMENT NO. 142 275

BASES FOR 3.6.12 AND 4.6.12 REACTOR PROTECTION SYSTEM AND REACTOR TRIP SYSTEM POWER SUPPLY MONITORING To eliminate the potential for undetectable single component failure which could adversely affect the operability of the reactor protection system and reactor trip system, protective relaying schemes are installed on Motor Generator Sets 131 and 141, Static Uninterruptible Power Supply Systems 162 and 172, and maintenance bus 130A. This provides for overvoltage, undervoltage and underfrequency protection.

AMENDMENT NO. 142 276

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4

3.6.13 REMOTE SHUTDOWN PANELS Applies to the operating status of the remote shutdown panels.

Oblective:

To assure the capability of the remote shutdown panels to provide 1) initiation of the emergency condensers Independent of the main/auxiliary control room 2) control of the motor-operated steam supply valves Independent of the main/auxiliary control room and 3) parameter monitoring outside the control room.

Soecifcation:

a.

During power operation, the remote shutdown panels' Functions in Table 3.6.13-1 shall be operable.

4.6.13 REMOTE SHUTDOWN PANELS Applicability:

Applies to the periodic testing requirements for the remote shutdown panels.

Obiective:

To assure the capability of the remote shutdown panels to provide 1) initiation of the emergency condensers independent of the main/auxiliary control room 2) control of the motor-operated steam supply valves independent of the main/auxiliary control room and 3) parameter monitoring outside the control room.

Specification:

The remote shutdown panels surveillance shall be performed as indicated below:

a.

Each remote shutdown panel monitoring instrumentation channel shall be demonstrated operable by performance of the operations and bshwn in Table 4.6.13 1.

1.

Each remote shutdown panel shall be demonstrated to initiate the emergency condensers independent of the main/auxiliary control room.

at hirqec speci4L d tu th AMENDMENT NO. 4.4, 155 277

TABLE 4.6.13-1 REMOTE SHUTDOWN PANEL MONITORING Parameter Reactor Pressure Reactor Water Level Reactor Water Temperature Torus Water Temperature Drywall Pressure Emergency Condenser Water Level Drywall Temperature "All Rods In" Light IkrwL7 F7 Sensor Check 2 Onee peF dy Gnee PeF day Gnee peF dy Gnee peF day Gnee peF 4.0%

Onee peF ft Onee pFd tkmuw 1) inft@roornanf V'henngsI fPaikra*jl Onee peF 3 FMeR1(a)

O.ne peF 3 FMeP, tt*()

Onoc per rofuoeling oyelc On.c p*c refu.ling y,*-

Once per 3 fuo!ngya)

Onoc pcr rofueling

/yolc N/A FIRA ar FRI No F1 a wo r---

(a)

The Indicator located at the remote shutdown panel will be calibrated at the frequency listed in Table 4.6.13-1. Calibration of the remaining channel Instrumentation is provided by Specification 4.6.2.

AMENDMENT NO. 142 280

BASES FOR 3.6.13 AND 4.6.13 REMOTE SHUTDOWN PANELS The remote shutdown panels provide 1) manual initiation of the emergency condensers 2) manual control of the steam supply valves and 3) parameters monitoring independent of the main/auxiliary control room. Two panels are provided, each located in a separate fire area, for added redundancy. Both panels are also in separate fire areas from the main/auxiliary control room. One channel of each Function provides the necessary capabilities consistent with 10 CFR 50.48(c). Therefore, only one channel of either remote shutdown panel monitoring instrument or control is required to be operable. The electrical design of the panels is such that no single fire can cause loss of both emergency condensers.

Each remote shutdown panel is provided with controls for one emergency condenser loop. The emergency condensers are designed such that automatic initiation is independently assured in the event of a fire 1) in the Reactor Building (principle relay logic located in the auxiliary control room or 2) in the main/auxiliary control room or Turbine Building (redundant relay logic located in the Reactor Building). Each remote shutdown panel also has controls to operate the two motor-operated steam supply valves on its respective emergency condenser loop. A key operated bypass switch is provided to override the automatic isolation signal to these valves. Once the bypass switch is activated, the steam supply valves can be manually controlled from the remote shutdown panels. Since automatic initiation of the emergency condenser is assured, the remote shutdown panels serve as additional manual controlling stations for the emergency condensers. In addition, certain parameters are monitored at each remote shutdown panel.

The remote shutdown panels are normally de-energized, except for the monitoring instrumentation, which is normally energized. To energize the remaining functions on a remote shutdown panel, a power switch located on each panel must be activated. Once the panels are completely energized, the emergency condenser condensate return valve and steam supply valve controls can be utilized.

AMENDMENT NO. !42, 155, Revision 37 (A215) 281

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 1 3.6.15 MAIN CONDENSER OFFGAS 4.6.15 MAIN CONDENSER OFFGAS Aoolicabilitv:

Applies to the radioactive effluents from the main condenser.

Applies to the periodic test and requirements of main condenser offgas.

recording To assure that radioactive material is not released to the environment in any uncontrolled manner and is within the limits of 1 OCFR20 and 1 OCFR50, Appendix I.

The gross radioactivity (beta and/or gamma) rate of noble gases measured at the recombiner discharge shall be limited to less than or equal to 500,000 pCI/sec.

This limit can be raised to 1 Cl/sec. for a period not to exceed 60 days provided the offgas treatment system is in operation.

With the gross radioactivity (beta and/or gamma) rate of noble gases at the recombiner discharge exceeding the above limits, restore the gross radioactivity rate to within its limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-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 />.

Obiective:

To ascertain that radioactive effluents from the main condenser are within allowable values of 10CFR20, Appendix B and 10CFR50, Appendix I.

Specification:

The gross radioactivity (beta and/or gamma) rate of noble gases from the recombiner discharge shall be determined to be within the limits of Specification 3.6.15 at-th fellwing frequ*ic*,

by performing an isotopic analysis of a representative sample of gases taken at the recombiner discharge:

Within 4

hours following an increase on the recombiner discharge monitor of greater than 50%, factoring out increases due to changes in thermal power level and dilution flow changes.

AMENDMENT NO. 442, 176 295

BASES FOR 3.6.15 AND 4.6.15 MAIN CONDENSER OFFGAS Restricting the gross radioactivity rate of noble gases from the main condenser provides assurance that the total effective dose equivalent to an Individual at the exclusion area boundary will not exceed a very small fraction of the limits of 10 CFR 50.67 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

The primary purpose of providing this specification is to limit buildup of fission product activity within the station systems which would result if high fuel leakage were to be permitted over extended periods.

.A.M_.B..1-E.ENT NO. 142, 176, Revision 21 (A194) 296

I IUITIKI tfL1r%1T1fL1 C 0 ACCOATIMI Q1 1PVF11 I AMtP 12Fn"10PUPKIT I

IUTIM~ P~k~fIT~~M AD A~DAIAMQIID~III A1~

DflhIDFFNI 3.7.1 SPECIAL TEST EXCEPTION - SHUTDOWN MARGIN DEMONSTRATIONS 4.7.1 SPECIAL TEST EXCEPTION - SHUTDOWN MARGIN DEMONSTRATIONS Applies to shutdown margin demonstration in the cold shutdown condition.

Oblective:

To assure the capability of the control rod system to control core reactivity.

a.

The reactor mode switch may be placed in the startup position to allow more than one control rod to be withdrawn for shutdown margin demonstration, provided that at least the following requirements are satisfied.

(1) The source range monitors are operable in the noncoincident condition.

(2) The rod worth minimizer is operable per Specification 3. 1.1 b(3)(b) and is programmed for the shutdown margin demonstration, or conformance with the shutdown margin demonstration procedure is verified by a second licensed operator or other technically qualified member of the unit technical staff.

AMENDMENT NO. 142 Applicability:

Applies to periodic inspections required to perform shutdown margin demonstrations in the cold shutdown condition.

Obiective:

To specify the inspections required to perform the shutdown margin demonstration in the cold shutdowr condition.

a.

Within 30 minutes prior to anda 12 h'euis during the performance of a shutdown margin demonstration, verify that:

(1) The source range monitors are operable per Specification 3.5.1.

(2) The rod worth minimizer is operable with the required program per Specification 3.1. lb(3)(b) or a second licensed operator or other technically qualified member of the unit technical staff is present and verifies compliance with the shutdown margin demonstration procedure.

339

BASES FOR 3.7.1 AND 4.7.1 SHUTDOWN MARGIN DEMONSTRATION The shutdown margin demonstration may be performed prior to power operation.

However, the mode switch must be placed in the startup I position to allow withdrawal of more than one control rod. Specifications 3.7.1 and 4.7.1 require certain restrictions in order to ensure that an Inadvertent criticality does not occur while performing the shutdown margin demonstration.

This special test exception provides the appropriate additional controls to allow the shutdown margin demonstration to be performed in the cold shutdown condition with the vessel head in place.

Compliance with this special test exception is optional and applies only if the shutdown margin demonstration will be performed prior to the reactor coolant system pressure and control rod scram time tests following refueling outages when core alterations are performed. The shutdown margin demonstration is performed using the in-sequence non-critical method.

A.MEN-DMENT NO. 1 1.2, RevisIon 5 (A180) 341

e.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the 10 CFR 50 Appendix J Testing Program Plan.

The provisions of Specification 4.0.3 are applicable to the 10 CFR 50 Appendix J Testing Program Plan.

6.5.8 Control Room Envelope Habitability Proaram A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Air Treatment (CRAT) System, CRE occupants can control the reactor safely under normal conditions and maintain It in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (I) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation of the CRAT System, operating at a flow rate of 2025-2475 cfm, at a Frequency of 24 months. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges Is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of TS 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

AMENDMENT NO. 4-.9, 202 355a

ATTACHMENT 4 License Amendment Request Nine Mile Point Nuclear Station, Unit 1 Docket No. 50-220 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

TSTF-425 (NUREG-1433) vs. NMP Unit 1 Cross-Reference

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 1 of 11 TSTF-425 (NUREG-1433) vs. NMP Unit 1 Cross-Reference Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Control Rod Operability 3.1.3 3.1.1 Control rod position 3.1.3.1 Notch test - fully withdrawn control rod one notch 3.1.3.2 4.1.1.(2)

Notch test - partially withdrawn control rod one notch 3.1.3.3 4.1.1.(2)

Control Rod Scram Times 3.1.4 Scram time testing 3.1.4.2 4.1.i.c(3)

Control Rod Scram Accumulators 3.1.5 Control rod scram accumulator pressure 3.1.5.1 4.1.1.d Rod Pattern Control 3.1.6

[BPWS] Analyzed rod position sequence 3.1.6.1 Standby Liquid Control (SLC) System 3.1.7 3.1.2 Volume of sodium pentaborate 3.1.7.1 4.1.2.b.2 Temperature of sodium pentaborate solution 3.1.7.2 4.1.2.b.3 Temperature of pump suction piping 3.1.7.3 Continuity of explosive charge 3.1.7.4 4.1.2.a.1 Concentration of boron solution 3.1.7.5 4.1.2.b. 1 Manual/power operated valve position 3.1.7.6 Pump flow rate 3.1.7.7 4.1.2.a.2 Flow through one SLC subsystem 3.1.7.8 4.1.2.a.1 Heat traced piping is unblocked 3.1.7.9 Verify sodium pentaborate enrichment [Solution Boron-10 3.1.7.10 4.1.2.b.4 Enrichment] (NUREG-1433 - prior to addition to SLC tank)

Surveillances with inoperable components 4.1.2.c Scram Discharge Volume (SDV) Vent & Drain Valves 3.1.8 Each SDV vent & drain valve open 3.1.8.1 4.1.1.e(1)

Cycle each SDV vent & drain valve fully closed/fully open position 3.1.8.2 4.1.1.e(2)

Each SDV vent & drain valve closes on receipt of scram 3.1.8.3 4.1.1.e(1)

Average Planar Linear Heat Generation Rate (APLHGR) 3.2.1 APLHGR less than or equal to limits 3.2.1.1 4.1.7.a Minimum Critical Power Ratio (MCPR) 3.2.2 MCPR greater than or equal to limits 3.2.2.1 4.1.7.c Linear Heat Generation Rate (LHGR) 3.2.3 LHGR less than or equal to limits 3.2.3.1 4.1.7.b Power Flow Relationship 4.1.7.d Average Power Range Monitor (APRM) Gain & Setpoints 3.2.4 MFLPD is within limits 3.2.4.1 4.6.2.c APRM setpoints or gain are adjusted for calculated MFLPD 3.2.4.2 4.6.2.c Reactor Protection System (RPS) Instrumentation 3.3.1.1 3.6.2 Channel Check 3.3.1.1.1 Per Table 4.6.2.a Absolute diff. between APRM channels & calculated power 3.3.1.1.2 4.6.2.c Adjust channel to conform to calibrated flow (APRM STP - Hi) 3.3.1.1.3 Table 4.6.2.a Channel Functional Test (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering Mode 2) 3.3.1.1.4 Channel Functional Test (weekly/monthly) 3.3.1.1.5

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 2 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Calibrate local power range monitors 3.3.1.1.6 Table 4.6.2.a Channel Functional Test 3.3.1.1.7 Calibrate trip units (quarterly) 3.3.1.1.8 Channel Calibration (APRMs) 3.3.1.1.9 Table 4.6.2.a Channel Functional Test (Reactor Mode Switch) 3.3.1.1.10 Channel Calibration 3.3.1.1.11 Per Table 4.6.2.a Verify APRM Flow Biased STP - High 3.3.1.1.12 Table 4.6.2.a Logic System Functional Test 3.3.1.1.13 Verify TSV/TCV closure/Trip Oil Press-Low Not Bypassed 3.3.1.1.14 Table 4.6.2.a Verify RPS Response Time 3.3.1.1.15 Source Range Monitor (SRM) 3.3.1.2 Channel Check 3.3.1.2.1 Verify Operable SRM Detector 3.3.1.2.2 Channel Check 3.3.1.2.3 Verify count rate 3.3.1.2.4 Channel Functional Test (7 days) 3.3.1.2.5 Channel Functional Test (31 days) 3.3.1.2.6 Channel Calibration 3.3.1.2.7 Control Rod Block Instrumentation 3.3.2.1 3.6.2 Channel Functional Test (routine) 3.3.2.1.1 Channel Functional Test (rod withdrawal at < 10% RTP) 3.3.2.1.2 Channel Functional Test (thermal power < 10%)

3.3.2.1.3 Verify RBM 3.3.2.1.4 Verify RWM not bypassed 3.3.2.1.5 Channel Functional Test 3.3.2.1.6 Table 4.6.2g Channel Calibration 3.3.2.1.7 Feedwater & Main Turbine High Water Level Trip Instrumentation 3.3.2.2 Channel Check 3.3.2.2.1 Channel Functional Test 3.3.2.2.2 Channel Calibration 3.3.2.2.3 Logic System Functional Test 3.3.2.2.4 Post Accident Monitor (PAM) Instrumentation 3.3.3.1 3.6.11 Channel Check 3.3.3.1.1 Per Table 4.6.11 Calibration 3.3.3.1.2 Per Table 4.6.11 Remote Shutdown System [Alternate Shutdown Monitoring]

3.3.3.2 3.6.13 Channel Check 3.3.3.2.1 Per Table 4.6.13-1 Verify control circuit and transfer switch capable of function 3.3.3.2.2 Channel Calibration 3.3.3.2.3 Per Table 4.6.13 -1 End-of-Cycle-Recirculation Pump Trip (RPT) Instrumentation 3.3.4.1 Channel Functional Test 3.3.4.1.1 Calibrate trip units 3.3.4.1.2 Channel Calibration 3.3.4.1.3

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 3 of 11 Technical Specification Section Title/Surveillance Descriptoon*

TSTF-425 NMP-1 Logic System Functional Test 3.3.4.1.4 Verify TSV/TCV Closure/Trip Oil Press-Low Not Bypassed 3.3.4.1.5 Verify EOC-RPT System Response Time 3.3.4.1.6 Determine RPT breaker interruption time 3.3.4.1.7 Anticipated Trip Without Scram-RPT Instrumentation 3.3.4.2 Channel Check 3.3.4.2.1 Channel Functional Test 3.3.4.2.2 Calibrate trip units 3.3.4.2.3 Channel Calibration 3.3.4.2.4 Logic System Functional Test 3.3.4.2.5 Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.1 Channel Check 3.3.5.1.1 Tables 4.6.2.d and f Channel Functional Test 3.3.5.1.2 Calibrate trip units 3.3.5.1.3 Channel Calibration (HPCI: Condensate Storage Tank Level - Low) 3.3.5.1.4 Channel Calibration 3.3.5.1.5 Tables 4.6.2.d and f Logic System Functional Test 3.3.5.1.6 Verify ECCS Response Time 3.3.5.1.7 Reactor Core Isolation Cooling (RCIC) System Instrumentation 3.3.5.2 Table 4.6.2.c

[Isolation Condenser Instrumentation]

Channel Check 3.3.5.2.1 Table 4.6.2.c Channel Functional Test 3.3.5.2.2 Calibrate trip units 3.3.5.2.3 Channel Calibration (Condensate Storage Tank Level - Low) 3.3.5.2.4 Channel Calibration 3.3.5.2.5 Table 4.6.2.c Logic System Functional Test 3.3.5.2.6 Primary Containment Isolation Instrumentation 3.3.6.1 Table 4.6.2.b Channel Check 3.3.6.1.1 Table 4.6.2.b Channel Functional Test 3.3.6.1.2 Calibrate trip units 3.3.6.1.3 Channel Calibration 3.3.6.1.4 Table 4.6.2.b Channel Functional Test (HPCI/RCIC Suppr. Pool Area Temp.)

3.3.6.1.5 Channel Calibration 3.3.6.1.6 Logic System Functional Test 3.3.6.1.7 Verify Isolation Response Time 3.3.6.1.8 Secondary Containment Isolation Instrumentation 3.3.6.2 Channel Check 3.3.6.2.1 Channel Functional Test 3.3.6.2.2 Calibrate trip units 3.3.6.2.3 Channel Calibration (Refueling Floor Exhaust Rad. - High) 3.3.6.2.4 Channel Calibration 3.3.6.2.5 Logic System Functional Test 3.3.6.2.6 Verify Isolation Response Time 3.3.6.2.7

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 4 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Low-Low-Set (LLS) Instrumentation 3.3.6.3 Channel Check 3.3.6.3.1 Channel Functional Test 3.3.6.3.2 Channel Functional Test 3.3.6.3.3 Channel Functional Test 3.3.6.3.4 Calibrate trip units 3.3.6.3.5 Channel Calibration 3.3.6.3.6 Logic System Functional Test 3.3.6.3.7 Main Control Room Environmental Control (MCREC) I 3.3.7.1 Table 4.6.2j Emergency Ventilation Initiation Channel Check 3.3.7.1.1 Channel Functional Test 3.3.7.1.2 Calibrate trip units 3.3.7.1.3 Channel Calibration 3.3.7.1.4 Logic System Functional Test 3.3.7.1.5 Loss of Power (LOP) Instrumentation / Diesel Generator Initiation 3.3.8.1 Table 4.6.2 i Channel Check 3.3.8.1.1 Channel Functional Test 3.3.8.1.2 Channel Calibration 3.3.8.1.3 Logic System Functional Test 3.3.8.1.4 RPS Electric Power Monitoring 3.3.8.2 3.6.12 Channel Functional Test 3.3.8.2.1 4.6.12.a Channel Calibration (RPS MG set/alt. power supply monitoring) 3.3.8.2.2 4.6.12.b System functional test 3.3.8.2.3 4.6.12.b Recirculation Loops Operating 3.4.1 Recirc loop jet pump flow mismatch with both loops operating 3.4.1.1 Jet Pumps 3.4.2 Criteria satisfied for each operating recirc loop 3.4.2.1 Safety/Relief Valves (SRVs) 3.4.3 3.2.9 Safety function lift setpoints 3.4.3.1 4.2.9.c SRV opens when manually actuated 3.4.3.2 4.2.9.b Reactor Coolant System (RCS) Operational Leakage 3.4.4 3.2.5 RCS unidentified and total leakage increase within limits 3.4.4.1 4.2.5.a RCS Pressure Isolation Valve (PIV) Leakage 3.4.5 Equivalent leakage of each PIV 3.4.5.1 RCS Leakage Detection Instrumentation 3.4.6 3.2.5 Channel Check 3.4.6.1 Channel Functional Test 3.4.6.2 4.2.5.b(2)

Channel Calibration 3.4.6.3 4.2.5.b(1)

RCS Specific Activity 3.4.7 3.2.4 Dose Equivalent 1-131 specific activity 3.4.7.1 4.2.4.a RCS Chemistry 3.2.3 Chloride ion content and conductivity 4.2.3 Residual Heat Removal (RHR) Shutdown Cooling - Hot Shutdown 3.4.8 One RHR Shutdown cooling subsystem operating 3.4.8.1 RHR Shutdown Cooling - Cold Shutdown 3.4.9

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 5 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 One RHR Shutdown cooling subsystem operating 3.4.9.1 RCS Pressure/Temperature Limit 3.4.10 RCS pressure, temperature, heatup and cooldown rates 3.4.10.1 RPV flange/head flange temperatures (tensioning head bolt stud) 3.4.10.7 RPV flange/head flange temperatures (after RCS temp < 800F) 3.4.10.8 RPV flange/head flange temperatures (after RCS temp < 1000F) 3.4.10.9 Reactor Steam Dome Pressure 3.4.11 Verify reactor steam dome pressure 3.4.11.1 ECCS - Operating / Core Spray and ADS 3.5.1 3.1.4 & 3.1.5 Verify injection/spray piping filled with water 3.5.1.1 Verify each valve in flow path is in correct position 3.5.1.2 Verify ADS nitrogen pressure 3.5.1.3 Verify RHR (LPCI) cross tie valve is closed and power removed 3.5.1.4 Verify LPCI inverter output voltage 3.5.1.5 Verify ECCS pumps develop specified flow 3.5.1.7 4.1.4.b 4.1.6.a 4.1.6.b Verify HPCI flow rate (Rx press < [1020], > [920])

3.5.1.8 Verify HPCI flow rate (Rx press < [165])

3.5.1.9 Verify ECCS actuates on initiation signal 3.5.1.10 Verify ADS actuates on initiation signal 3.5.1.11 4.1.5.b Verify each ADS valve opens [actuator strokes] when manually 3.5.1.12 4.1.5.a actuated Verify motor-operated valve operability 4.1.4.c

[Core Spray System, Containment Cooling System, Fire Protection System]

Verify pump compartment water-tight doors closed 4.4.3.a

[Core Spray System, Containment Cooling System]

Verify injection/spray piping filled with water 4.1.4.g Surveillance with inoperable System 4.1.4.e 4.1.6.c ECCS - Shutdown 3.5.2 Verify, for LPCI, suppression pool water level 3.5.2.1 4.3.2.a Verify, for CS, suppression pool water level and CST water level 3.5.2.2a 4.1.4.f 3.5.2.2b Verify ECCS piping filled with water 3.5.2.3 4.1.4.g Verify each valve in flow path is in correct position 3.5.2.4 Verify each ECCS pump develops flow 3.5.2.5 4.1.4.b Verify ECCS actuates on initiation signal 3.5.2.6 RCIC System [Isolation Condenser]

3.5.3 Verify RCIC piping filled with water 3.5.3.1 Verify each valve in flow path is in correct position 3.5.3.2 Verify RCIC flow rate (Rx press < [1020], > [920])

3.5.3.3 Verify RCIC flow rate (Rx press < [165])

3.5.3.4 Verify RCIC actuates on initiation signal 3.5.3.5 Emergency Cooling System 3.1.3 System Heat Removal Capability 4.1.3.

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 6 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Shell Side and Makeup Tank water level 4.1.3.b Manual open/close makeup tank water level control valve 4.1.3.c Area Temperature 4.1.3.d Surveillance with inoperable System 4.1.3.f Primary Containment 3.6.1.1 3.3.0 Verify drywell to suppression chamber differential pressure 3.6.1.1.2 Primary Containment Air Lock 3.6.1.2 Verify only one door can be opened at a time 3.6.1.2.2 Primary Containment Isolation Valves (PCIVs) 3.6.1.3 3.3.4 Verify purge valve is closed except one valve in a penetration 3.6.1.3.1 Verify each 18 inch (6 inch & 18 inch) PC purge valve is closed 3.6.1.3.2 Verify each manual PCIV outside containment is closed 3.6.1.3.3 Verify continuity of traversing incore probe (TIP) shear valve 3.6.1.3.5 Verify isolation time of each power operated PCIV 3.6.1.3.6 4.3.4.a Perform leakage rate testing on each PC purge valve 3.6.1.3.7 Verify isolation time of MSIVs 3.6.1.3.8 Verify automatic PCIV actuates to isolation position 3.6.1.3.9 4.3.4.a Cycle normally open isolation MOVs 4.3.4.b Verify sample of Excess Flow Check Valves actuate to isolation 3.6.1.3.10 4.3.4.c position Test explosive squib from each shear valve 3.6.1.3.11 Verify each purge valve is blocked 3.6.1.3.15 Drywell Pressure 3.6.1.4 Verify drywell pressure is within limit 3.6.1.4.1 Drywell Average Air Temperature 3.6.1.5 Verify drywell average air temperature is within limit 3.6.1.5.1 LLS Valves 3.6.1.6 Verify each LLS valve opens when manually actuated 3.6.1.6.1 Verify LLS system actuates on initiation signal 3.6.1.6.2 Reactor Building - Suppression Chamber Vacuum Breakers 3.6.1.7 3.3.6 Verify each vacuum breaker is closed 3.6.1.7.1 Perform functional test on each vacuum breaker 3.6.1.7.2 4.3.6.c(1)

Verify opening setpoint for each vacuum breaker 3.6.1.7.3 4.3.6.c(2)

Air-operated vacuum breaker instrumentation calibration Suppression Chamber - Drywell Vacuum Breakers 3.6.1.8 3.3.6 Verify each vacuum breaker is closed 3.6.1.8.1 Perform functional test on each vacuum breaker 3.6.1.8.2 4.3.6.a Verify opening setpoint for each vacuum breaker 3.6.1.8.3 Vacuum breaker position indication and alarm calibration Vacuum breaker inspection 4.3.6.b(3)

Main Steam Isolation Valve (MSIV) Leakage Control System 3.6.1.9 N/A Operate each MSIV LCS blower 3.6.1.9.1 Verify continuity of inboard MSIV LCS heater element 3.6.1.9.2 Perform functional test of each MSIV LCS subsystem 3.6.1.9.3 Suppression Pool Average Temperature 3.6.2.1 3.3.2

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 7 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Verify suppression pool average temperature within limits 3.6.2.1.1 4.3.2.a Suppression Pool Water Level 3.6.2.2 3.3.2 Verify suppression pool water level within limits 3.6.2.2.1 4.3.2.a Suppression Chamber Level, Temperature and Pressure 3.3.2 Suppression chamber temperature and pressure 4.3.2.a Suppression chamber visual inspection 4.3.2.b RHR Suppression Pool Cooling 3.6.2.3 Verify each valve in flow path is in correct position 3.6.2.3.1 Verify each RHR pump develops flow rate 3.6.2.3.2 RHR Suppression Pool Spray 3.6.2.4 Verify each valve in flow path is in correct position 3.6.2.4.1 Verify RHR pump develops flow rate 3.6.2.4.2 4.3.7.a.2 Drywell - Suppression Chamber Differential Pressure 3.6.2.5 Verify differential pressure is within limit 3.6.2.5.1 Drywell Cooling System Fans 3.6.3.1 Operate each fan > 15 minutes 3.6.3.1.1 Verify each fan flow rate 3.6.3.1.2 Primary Containment Oxygen Concentration 3.6.3.2 3.3.1 Verify PC oxygen concentration is within limits 3.6.3.2.1 4.3.1 Containment Atmosphere Dilution (CAD) System 3.6.3.3 Verify CAD liquid nitrogen storage 3.6.3.3.1 Verify each CAD valve in flow path is in correct position 3.6.3.3.2 Secondary Containment 3.6.4.1 3.4 Verify SC vacuum is > 0.25 inch of vacuum water gauge 3.6.4.1.1 4.4.1 Verify all SC equipment hatches closed and sealed 3.6.4.1.2 Verify one SC access door in each opening is closed 3.6.4.1.3 4.4.3.b(1)

Verify SC drawn down using one SGTS 3.6.4.1.4 4.4.1 Verify SC can be maintained using one SGTS 3.6.4.1.5 4.4.1 Core and Containment Spray Compartments 4.4.3.a Secondary Containment Isolation Valves / Rector Building 3.6.4.2 3.4.2 Integrity - Isolation Valves Verify each SC isolation manual valve is closed 3.6.4.2.1 Verify isolation time of each SCIV 3.6.4.2.2 Verify each automatic SCIV actuates to isolation position 3.6.4.2.3 4.4.2 Standby Gas Treatment (SGT) System 3.6.4.3 Operate each SGT subsystem for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 3.6.4.3.1 SGT filter testing [NUREG-1433 - Ventilation Filter Testing 3.6.4.3.2 Program]

Verify each SGT subsystem actuates on initiation signal 3.6.4.3.3 Verify each SGT filter cooler bypass damper can be opened 3.6.4.3.4 Residual Heat Removal Service Water (RHRSW) System 3.7.1 Verify each RHRSW valve in flow path in correct position 3.7.1.1 Plant Service Water (PSW) System and Ultimate Heat Sink (UHS) 3.7.2 Verify water level in cooling tower basin 3.7.2.1 Verify water level in pump well of pump structure 3.7.2.2 Verify average water temperature of heat sink 3.7.2.3 4.3.7.f

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 8 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Operate each cooling tower fan 3.7.2.4 Verify each PSW SSW valve in flow path is in correct position 3.7.2.5 Verify PSW SSW actuates on initiation signal 3.7.2.6 Diesel Generator (DG) Standby Service Water (SSW) System 3.7.3 Verify each DG SSW valve in flow path is in correct position 3.7.3.1 Verify DG SSW pump starts automatically when DG energizes the 3.7.3.2 4.6.3.b respective bus MCREC [Control Room HVAC] System / Control Room Air 3.7.4 Treatment System Operate each MCREC [Control Room HVAC] subsystem 3.7.4.1 4.4.5.e Test and Sampling analyses 4.4.5.b Verify each subsystem actuates on initiation signal 3.7.4.3 4.4.5.f Verify each subsystem can maintain positive pressure 3.7.4.4 Verify DP across HEPA filters 4.4.5.a Control Room Air Conditioning System 3.7.5 Verify each subsystem has capability to remove heat load 3.7.5.1 Emergency Ventilation System 3.4.4 Pressure drop across HEPA and Charcoal adsorbers 4.4.4.a(1)

Test and sample analysis 4.4.4.b Each circuit shall be operated 4.4.4.e Automatic initiation of each branch 4.4.4.g Operability of bypass valve for filter cooling 4.4.4.h Surveillances with inoperable components 4.4.4.i Main Condenser Offgas 3.7.6 3.6.15 Verify gross gamma activity rate of the noble gases 3.7.6.1 4.6.15 Main Turbine Bypass System 3.7.7 Verify one complete cycle of each main turbine bypass valve 3.7.7.1 Perform system functional test 3.7.7.2 Verify Turbine Bypass System Response Time within limits 3.7.7.3 Spent Fuel Storage Pool Water Level 3.7.8 Verify spent fuel storage pool water level 3.7.8.1 AC Sources - Operating 3.8.1 3.6.3 Verify correct breaker alignment 3.8.1.1 Verify each DG starts from standby conditions/steady state 3.8.1.2 4.6.3.b Verify each DG is synchronized and loaded 3.8.1.3 4.6.3.b Verify each day tank level 3.8.1.4 Check for and remove accumulated water from day tank 3.8.1.5 Verify fuel oil transfer system operates 3.8.1.6 Verify each DG starts from standby conditions/quick start 3.8.1.7 4.6.3.b Verify transfer of power from offsite circuit to alternate circuit 3.8.1.8 Verify DG rejects load greater than single largest load 3.8.1.9 Verify DG maintains load following load reject 3.8.1.10 Verify on loss of offsite power signal 3.8.1.11 4.6.3.a Verify DG starts on ECCS initiation signal 3.8.1.12 4.6.3.a Verify DG automatic trips bypassed on ECCS initiation signal 3.8.1.13 Verify each DG operates for > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.8.1.14

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 9 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Verify each DG starts from standby conditions/quick restart 3.8.1.15 Verify each DG synchronizes with offsite power 3.8.1.16 Verify ECCS initiation signal overrides test mode 3.8.1.17 Verify interval between each timed load block 3.8.1.18 Verify on LOOP in conjunction with ECCS initiation signal 3.8.1.19 4.6.3.a Verify simultaneous DG starts 3.8.1.20 Diesel Generator Starting Batteries Verify electrolyte level, pilot cell voltage, battery voltage on float charge, pilot cell specific gravity Specific gravity and electrolyte temperature for each fourth cell Specific gravity and electrolyte temperature for every cell Battery capacity during battery service test Battery capacity during battery discharge test Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 Verify fuel oil storage tank volume 3.8.3.1 Verify lube oil inventory 3.8.3.2 Verify each DG air start receiver pressure 3.8.3.4 Check/remove accumulated water from fuel oil storage tank 3.8.3.5 DC Sources - Operating/Battery Parameters 3.8.4 Verify battery terminal voltage 3.8.4.1 4.6.3.c Verify each battery charger supplies amperage 3.8.4.2 Verify battery capacity during battery service test 3.8.4.3 Battery Parameters 3.8.6 Verify battery float current 3.8.6.1 Verify battery pilot cell voltage 3.8.6.2 4.6.3.c Verify battery connected cell electrolyte level 3.8.6.3 Verify battery pilot cell temperature 3.8.6.4 Verify battery connected cell voltage 3.8.6.5 Verify battery capacity during performance discharge test 3.8.6.6 Inverters - Operating 3.8.7 Verify correct inverter voltage, frequency and alignment 3.8.7.1 Inverters - Shutdown 3.8.8 Verify correct inverter voltage, frequency and alignment 3.8.8.1 Distribution System - Operating 3.8.9 Verify correct breaker alignment/power to distribution subsystems 3.8.9.1 Distribution System - Shutdown 3.8.10 Verify correct breaker alignment/power to distribution subsystems 3.8.10.1 Refueling Equipment Interlocks 3.9.1 3.5.2 Channel Functional Test of refueling equip interlock inputs 3.9.1.1 4.5.2 SRMs (Refueling) 3.5.1 SRMs checked, tested, and calibrated 4.5.1 Refuel Position One-Rod-Out Interlock 3.9.2 Verify reactor mode switch locked in refuel position 3.9.2.1 Perform Channel Functional Test 3.9.2.2 Control Rod Position 3.9.3 Verify all control rods fully inserted 3.9.3.1

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 10 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Control Rod Operability - Refuel 3.9.5 Insert each withdrawn control rod one notch 3.9.5.1 Verify each withdrawn control rod scram accumulator press 3.9.5.2 Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel 3.9.6 Verify RPV water level 3.9.6.1 Reactor Pressure Vessel (RPV) Water Level - New Fuel 3.9.7 Verify RPV water level 3.9.7.1 RHR - High Water Level 3.9.8 Verify one RHR shutdown cooling subsystem operating 3.9.8.1 RHR - Low Water Level 3.9.9 Verify one RHR shutdown cooling subsystem operating 3.9.9.1 Reactor Mode Switch Interlock Testing 3.10.2 Verify all control rods fully inserted in core cells 3.10.2.1 Verify no core alterations in progress 3.10.2.2 Single Control Rod Withdrawal - Hot Shutdown 3.10.3 Verify all control rods in five-by-five array are disarmed 3.10.3.2 Verify all control rods other than withdrawn rod are fully inserted 3.10.3.3 Single Control Rod Withdrawal - Cold Shutdown 3.10.4 Verify all control rods in five-by-five array are disarmed 3.10.4.2 Verify all control rods other than withdrawn rod are fully inserted 3.10.4.3 Verify a control rod withdrawal block is inserted 3.10.4.4 Single Control Rod Drive (CRD) Removal - Refuel 3.10.5 Verify all control rods other than withdrawn rod are fully inserted 3.10.5.1 Verify all control rods in five-by-five array are disarmed 3.10.5.2 Verify a control rod withdrawal block is inserted 3.10.5.3 Verify no core alterations in progress 3.10.5.5 Multiple CRD Removal-Refuel 3.10.6 Verify four fuel assemblies removed from core cells 3.10.6.1 Verify all other rods in core cells inserted 3.10.6.2 Verify fuel assemblies being loaded comply with reload sequence 3.10.6.3 Shutdown Discharge Margin Test - Refueling 3.10.8 Verify no other core alterations in progress 3.10.8.4 Verify CRD charging water header pressure 3.10.8.6 Recirculation Loops - Testing 3.10.9 Verify LCO 3.4.1 requirements suspended for < 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.10.9.1 Verify Thermal power < 5% RTP during Physics Test 3.10.9.2 Training Startups 3.10.10 Verify all operable IRM channels are <25/40 div. of full scale 3.10.10.1 Verify average reactor coolant temperature < 200 F 3.10.10.2 Radioactive Liquid Storage Liquids contained in tanks sampled and analyzed Programs (Surveillance Frequency Control Program [SFCP])

5.5.15 6.5.9 High Pressure Core Injection 3.1.8 Pump auto start 4.1.8.a Pump operability 4.1.8.b Surveillances with inoperable components 4.1.8.c

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 11 of 11 Technical Specification Section Title/Surveillance Description*

TSTF-425 NMP-1 Radioactive Material Sources 3.6.5 Seal source leak test 4.6.5.1 Startup sources 4.6.5.3 Reactor Coolant System Isolation Valves 3.27 Automatic Initiation and Closure Times 4.2.7.a Feedwater and Main Steam partial closure and re-opening 4.2.7.c Pressure Relief Systems-Safety Valves 3.2.8 Remove and test for lift set point and partial test 4.2.8 Containment Spray System 3.3.7 Automatic Start of Containment Spray Pump 4.3.7.a(1)

Pump Operability 3.6.2.4.2 4.3.7.a(2)

Raw Water Cooling Pump Operability 4.3.7.c Surveillance with Inoperable equipment 4.3.7.d Lake water Temperature 3.7.2.3 4.3.7.f Instruments that Initiates Core Spray Per Table 4.6.2.d Instruments that Initiates Containment Spray Per Table 4.6.2.e Instrumentation that Initiates Control Rod Withdrawal Block Per Table 4.6.2.g Vacuum Pump Isolation Per Table 4.6.2.h Emergency Ventilation Initiation Per Table 4.6.2.j High Pressure Coolant Injection Per Table 4.6.2.k Control Room Air Treatment System Initiation Per Table 4.6.2.1 Mechanical Vacuum Pump Isolation 3.6.1 Automatic isolation of the mechanical vacuum pump 4.6.1.b Special Test Exception -SDM Demonstration 3.7.1 Operability of SRM and RWM 4.7.1.a The Technical Specification Section Title/Surveillance Description portion of this attachment is a summary description of the referenced TSTF-425 (NUREG-1433)/NMP-1 TS Surveillances which is provided for information purposes only and is not intended to be a verbatim description of the TS Surveillances.

ATTACHMENT 5 License Amendment Request Nine Mile Point Nuclear Station, Unit 1 Docket No. 50-220 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

Proposed No Significant Hazards Consideration

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 1 of 2 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request: This amendment request involves the adoption of approved changes to the standard technical specifications (STS) for General Electric Plants, BWR/4 (NUREG-1433), to allow relocation of specific TS surveillance frequencies to a licensee-controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).

The proposed changes are consistent with NRC-approved Industry/ TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b."

The proposed changes relocate surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456).

Basis for proposed no significant hazards consideration: As required by 10 CFR 50.91 (a),

the Exelon analysis of the issue of no significant hazards consideration is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program.

Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 2 of 2 changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Exelon will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP.

NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the above, Exelon concludes that the requested changes do not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), "Issuance of Amendment."

ATTACHMENT 6 License Amendment Request Nine Mile Point Nuclear Station, Unit 1 Docket No. 50-220 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

Proposed Inserts

LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-220 Page 1 of 1 INSERT 1 In [in] accordance with the Surveillance Frequency Control Program INSERT 2 6.5.9 Surveillance Frequency Control oroaram This program provides controls for the Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.
b. Changes to the Frequency listed in the Surveillance Frequency Controlled Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequency," Revision 1.
c. The provision of Surveillance Requirement 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

INSERT 3 The [the] Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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