ML20128M719

Proposed TS Increasing Minimum Core Spray Pump Flow to More Conservatively Account for ECCS Bypass Leakage Paths

ML20128M719
Person / Time
Site:	Monticello
Issue date:	02/12/1993
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20128M699	List: ML20128M695 ML20128M710 ML20128M719
References
NUDOCS 9302220255
Download: ML20128M719 (26)

Text

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4 4

Exhibit B Monticello Nuclear Generating Plant License Amendment Reauest Dated February 12. 1993 4

Technical Specification Pages Marked Up

with Proposed Wording Changes Exhibit B consists of the existing Technical Specification pages marked up with the proposed changes. Existing pages affected by this change are listed below

k EALt t

52 53 54 55 l 60d 101 107 110 113 127 15 '.

156 i

9302220255 930212 PDR ADOCK 05000263 p PDR

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Table 3.2.2 Instrumentation That Initiates Emergency Core Cooling Systems Minimum No. of Oper-Minimum No. of able or Operating Operable or Total No. of Instru- Instrument Chaniiels .

l Operating Trip ment Channels Per Per Trip System' Required .

Function Trio Settine Systems (3) Trio System (3) Conditions

• A. Core Sorav and LPCI

1. Pump Start

a. Low Low Reactor 26*6"56'10" 2 4(4) 4 A.

Water Level and

b. i. Reactor Low 2450 psig 2 2(4) 2 A.

Pressure Permissive or

11. Reactor Low 2011 min 2 1 1 GB.

Pressure Permissive-end ,

Bypass Timer i

c. High Drywell s2 psig 2 4(4) 4 A.

Pressure (1)

2. Low Reactor Pressure 2450 psig 2 2(4) 2 A.

(Valve Permissive)

3. Loss of Auxiliary - ----- 2 2(2) 2 A.

Power  ;

f i

1 3.2/4.2 52

Table 3.2.2 Instrumentation That Initiates Emergency Core Cooling Systems  ;

Minimum No. of Oper-Minimum No. of able or Operating

, Operable or Total No. of Instru- Instrument Channels Operating Trip ment Channels Per Per Trip System Required Function Trio Settinz Systems (3) Trio System (3) Conditions

• B. HPCI System

1. High Drywell-' 52 psig 1 4 4 RA.^~

Pressure (1)

2. Low-iow Reactor 26'6"s6'10" 1 4 4 RA' . ~

Water Level C. Automatic Depres-surization

1. Low-Lou Reactor 26'6"s6'10" 2 2 2 CB. ~

Water Level and

2. Auto Blowdown 5120 seconds 2 1 1 CB.

Timer and

3. Low Pressure Core $100 psig 2 12(4) 12(4) CB. i Cooling Pumps Dis-Charge Pressure Interlock i~ ,

e

. 3.2/4.2 53 1

i i

Table 3.2.2 - Continued Instrumentation That Initiates Emerzency Core Cooline System Min. No. of Oper-Min. No. able or Operating of Operable Total No. of Instru- Instrument Channels or Operating ment Channels Per Per Trip System Required Function Trio settine Trio Systems (3) Trio System (3) Conditions

• D. Diesel Generator

1. Degraded or Loss .

of Voltage Essential Bus (5)

2. Low Low Reactor 2:6'6"s6'10" 2 4(4) 4 DC.

~

Water Level

3. High Drywell Press 52 psig 2 4(4) 4 DC.

NOTES:

1. High drywell pressure may be bypassed when necessary only by closing the manual containment isolation valves during e purging for containment inerting or de-inerting. Verification of the bypass condition shall be noted in the control room log. Also.need not be operable when primary containment' integrity is not required.  ;

t

2. One instrument channel is a circuit breaker contact and the other is an undervoltage relay.

1 i

l; j 3.2/4.2. 54 I

l'

Table 3.2.2 - Continued Notes:

3. Upon discovery that minimum requirements for the number of operable or operating trip. systems, or instrument channels are not' satisfied action shall be initiated to:

(a) Satisfy the requirements by placing appropriate channels or systems in the tr'ipped condition, or (b) Place the plant under the'specified required conditions using normal operating procedures.

4. All instrument channels are shared by both trip systems.

5. See table 3.2.6.

• Required conditions when minimum conditions for operation are not satisfied.

A. Com; f with Specification 3.5.A.

A--rc= ply ri t' Specificctier 3.5.D.

GB. Reactor pressure 5150 psig.

DC. ' Comply with Specification 3.9.B.

I 3.2/4.2 55

4 Table 3.2.8 other Instrumentation Minimum No. of Minimum No. of Oper-Operable or Total No. of Instru- able or Operating . Required Function Trip Setting Operating Trip ment Channels Per Instrument Channels Conditions

• System (1) ' Trio System Per Trio System (1)

A. RCIC Initiation

1. Low-Low Reactor Level 26'6"& $6'10" 1  ? 2 B above top of '

i active fuel B. HPCI/RCIC Turbine Shutdown 4 a, liigh Reactor Level 514'6" above 1 2 2 A top of active 4

fuel C. HPCI/RCIC Turbine '

Suction Transfer >

a. Condensate Storage 22'0" above 1 2 2 C Tank Low Level tank bottom 1 t

NOTE:

1. Upon discovery that minimum requirements for the number of operable or operating trip systems or instrument channels are not satisfied, action shall be initiated to:

a. Satisfy the requirements by placing the . appropriate channels or systems in the tripped. condition 4

-(Turbine /Feedwater Trip only), or

b. Place-the plant under the specified required condition using normal operating procedures.

j

• Required. conditions when minimum conditions for operation are not satisfied:

A. Reactor in Startup, Refuel, or Shutdown Mode. ,

B. . Comply vith Specification 3.5.DFv3.

C. Align llPCI and RCIC suction to the suppression pool. Restore channels to operable status statu within 30 days j or place the plant _in Required Condition A.

i 3.2/4.2 60d i

i b

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT SPRAY / COOLING SYSTEMS 4.5 CORE AND CONTAINMENT SPRAY / COOLING SYSTEMS .

Applicability: Apolicability:

Applies to the. operational status of the emergency Applies to the periodic testing of the' emergency cooling systems. , cooling systems.

Obiective: Obiective:

To insure adequate cooling capability for heat removal To verify the operability of the emergency cooling in the event of a loss of coolant accident or systems.

Isolation from the normal reactor heat sink.

Specification: Specification: [

A. ECCS Systems A. ECCS Systems

1. Except as specified in section 3.5.A.3, both 1. Demonstrate the Core Spray Pumps develop a Core Spray subsystems and the Low Pressure 377CO;2J8QO gpm flow rate against a system Coolant Injection (LPCI) Subsystem (LPCI Mode head corresponding to a reactor pressure'of of RHR System) shall be operable whenever 130 psi greater than containment pressure, irradiated fuel is in the reactor vesset and. when tested pursuant to Specification the reactor water temperature is greater than 4.15.B.

212*F. . I

2. Demonstrate the LPCI Pumps develop a 3,870  !

2. Except as specified in section 3.5.A.3, the gpm flow rate against a system head

,High Pressure Coolant Injection (HPCI) System ' corresponding to two pumps delivering 7,740 and the Automatic Depressurization System gpm at a reactor pressure of 20 psi greater (ADS) shall be operable whenever the reactor than containment pressure, when tested pressure is greater than 150 psig and pursuant to Specification 4.15.B.

irradiated fuel is in the reactor vessel except during reactor vessel hydrostatic or

3. Demonstrate the HPCI Pump develops a 2700 1eakage tests. gpm flow rate against a reactor pressure

, range of 1120 psig to 150 psig, when tested pursuant to Specification 4.15.B.

s j 3.5/4.5 101 1

2

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

'F. Recirculation System F. Recirculation System

1. The reactor may'be started and operated, or 1. See Specification 4.6.c operation may. continue with only one recirculation loop in operation provided that: 2. The following baseline noise levels will be obtained prior to operation with only one

a. The following changes to setpoints and recirculation pump in operation at a core safety limit settings will be made within thermal power greater than that specified ,

in Figure 3.5.1 or with a core flow greater i

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiating operation with only one recirculation loop in operation. than 45% provided that baseline values have

' not been established since the last core

1. The Operating Limit MCPR (MCPR) will be refueling. Baseline values will be taken changed per Specification 3.11.C. with only one recirculation pump running.

2. The Maximum' Average Planar Linear Heat a. Establish a baseline core plate AP noise Generation Rate (MAPLHCR) will be level.

changedasnotedinTable1[6f[thg[ Cliff l 9kk h p M Q igity $epprt?.11.1. b. Establish a baseline APRM and LPRM ]'

neutron flux noise level.

3. The APRM Neutron Flux Scram and APRM Rod Block setpoints will be changed as noted 3. With only one recirculation loop in in Specification 2.3.A and-Table 3.2.3. operation at a core thermal power greater-than that specified in Figure 3.5.1 or with

b. Total core flow will be maintained greater a core flow greater than 45%, determine the than 39% when core thermal power is above following noise levels at least once per 8 the limit'specified in Figure 3.5.1. hour period and within 30 minutes after a core thermal power increase of greater than 5% of rated thermal power.

a. Core plate AP noise levels.

b. APRM and LPRM neutron noise levels.

107 3.5/4.5

. . . _ . - . - - . - - - -. .- . _ _ . -- -. - .. ~. - _ -

Bases 3.5/4.5 A. ECCS Systems The core spray system is provided to assure that the core is adequately cooled following a loss-of-coolant accident and, together with the LP,CI mode of the RHR system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the automatic depressurization system (ADS).

The Core Spray System is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining. The Core Spray pump is

' designed to deliver greater-than or equal to 3020 gpm (cafety cr:1yri  ::=^d 2700 gp-theTSAFER7CESTRILOCn AafstyTanslysis7sssumid?sTC6fd[ypHiy]iyilflWQdfG2@002^hpm*,p^G22003Kf)pagno]the icore} R100fgpm{to~

heepuntjfogECCS}bypas@ leakage) against a system head corresponding to a reactor pressure of 130 psi greater than containment pressure.

The surveillance requirements provide adequate assurance that the Core Spray System will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The low pressure coolant injection (LPCl) mode of the Fdut system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Four pumps are available to provide adequate core flooding for all break sizes'up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS. LPCI Loop Selection Logic determines which Recirculation loop the four RHR pumps will pump into. Each RHR pump was designed to deliver greater than or equal to 4000 gpm (the safety analysis assumed two pumps delivering 7,740 gpm) against a system head corresponding to a reactor pressure of 20 psi greater than containment pressure.

The allowed out-of-service conditions (Section 3.5.A.33) are determined from ECCS analysis cases analyzed.

Only one of these conditions is permitted to exist. If more than one condition exists, an orderly shutdown shall be initiated. A LPCI injection path consists of the two motor. operated injection valves on that path.

The surveillance _ requirements provide adequate assurance that the.LPCI system will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The 3.5/4.5 110

m. __. . .. .._ . ,__ _..~,_._-- __._. - _ ___ _. . . . . . . _ _ . . . . . ~ _ . _ - __ _ -.--_..e _ _ _ _ _

i 4

Bases 3 5/4.5 Continued:

The RHR service water system provides cooling for the RHR heat exchangers and can thus maintain the suppression pool water within limits. With the flow specified, the pool temperature limits are maintained as specified in Specification 3.7.A.l.

i D. RCIC TheRCICsystemisprovidedtosupp1Ycontinuousmakeupwatertothereactorcorewhenthereactorisolated from the turbine and when the feedwater system is not available. The pumping capacity of the RCIC system is sufficier.t to maintain the water level above the core without any other water system in operation. If the water level in the-reactor vessel decreases.to the RCIC initiation level, the system automatically starts.

The system may also be manually initiated at any time.

The HPCI system provides' an alternate method of supplying makeup water to the . reactor should the normal feedwater become unnvailable. Therefore, the specification calls for an operability check of the HPCI ,

system should the RCIC system be found to be inoperable.

- The surveillance requirements provide adequate assurance that the RCIC system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a '

test loop during. reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

E. Cold Shutdown and Refueling Requirements 1

The purpose of Specification 3.5.E is to assure .that sufficient core cooling equipment is available at all- ,

times. It is during refueling outages that major maintenance is performed and during such time that 4 all core and containment spray / cooling ' subsystems may be out of service. This specification allows all ,

core and containment spray / cooling subsystems to be inoperable provided no work is being done which has the f

, potential for draining the reactor vessel. Thus events requiring core cooling are precluded. -i 4

Specification 3.5.E.4[recognizesthatconcurrentwithcontrolroddrivemaintenanceduringthe  ;

refueling outage, it may by necessary to drain the suppression chamber.for maintenance or for.the i inspection required by Specification 4.7. A.l. In this situation, a sufficient inventory of- tater is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing i , or instrument thimble seal integrity.

i 1

1 3.5/4.5 Bases 113 J I

1 3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS i 7

E. Safety / Relief Valves E. Safety / Relief Valves  !

1. During power operating conditions and whenever 1. a. A minimum of seven safety / relief valves  ;

reactor coolant pressure is greater than 110 shall be bench checked or replaced with ,

psig and temperature is greater than 345'F ths a bench checked valve each refueling t asfetyfVsliF4?fsiistioiiff(iself ace 6sEidn)?6f ~ outage. The nominal self-actuation

• ssvennsfety/selikfWilvesGhd11tbeloperablu setpoints are specified in Section

"(nobeliLowsLoGlSetiand?ADSEreduirementsTare~ 2.4.B.

16Eated ? iniSpshifiAational322; H jand f3151Aj ~

Edspfstljelf)%~~~ ' ~ ~~ ~ ^ ~~~~~~ ~~ ' b. At least two of the safety / relief  ;

valves shall be disassembled and

. "": H fety elve functie (ccif inspected each refueling outage.  ;

cetuntien) of ::ve ecrety/ relief I

elve: ch:11 he per ble. c. The integrity of the safety / relief valve bellows shall be continuously 5 '" relencid netivated relief functier monitored. .!

(f.ute : tic Preccur: Relief} Fall be i eper ble 2: r: quired by Specificctier d. The operability of the bellows 2.5.E. monitoring system shall be demonstrated at least once every

c. TF L:v Lev S t. Functics fer three three months, ner

• uter tic Preccur: 9elief Velvec 2.

Fall he'eper:51: : requi cd by Low-Low Set Logic surveillance shall be

- b Sp::ificctier 2.2." performed in accordance with Table 4.2.1. 1

2. If Specification'3.6.E.1.re is not met, initiate an orderly shutdown and have' reactor '

coolant pressure and temperature reduced to 110 psig or less . and :345'F or less within 24 '  !

hours.-

1 t

.j

)

j. 3.~6/4.6 127 4

.i~

_. . _ _ _ . _ . . _. _ . ~ . _ . . _ . _ . _ . . . _ . . - - . _ _ _ _ . _ . _ _ _ . _ _ . - _ . _ . . _ _ _ _ _ - . _ _. . . . - - _ . _

Bases Continued 3.6 and 4.6: ,

t n e' car ty/relier valve: ' :: tu: functi:n ; i.e. p:rer relief er : 1f :tuated by 'igh pr ::ure.

5: ::1:ncid ::tuated functier (.*.ut: tic "rescur: " c l i e f) i rhi 5 r:ter ;l instrur--tari - -ign:12 mf

ircident high dry = 11 pre:curr end'lcr Ier rater 1:v 1 initict: epr-ing -f th: -217: Six functier i: -

discussed ir Specifientier 2.5.r !r raditic., th: :1v:: : r 5: :p; ret:d r u:117 m..-._-_._._.c.,

over-pressurede liefyd. m.l.f __. _dy b,y essure g ;.hi gh_j pr.-_.,._}y; e;sa rel ess aitwo.s ons N se -actuate 1 iib 6MDysydsdEza$$5/Phes$df$hdnNrdD[$d5(5LNTh n ws  : -* -m- mm fe_ty/. _ . .i_efRyal_v._,,h_av._

i re r .fu_nc. ti -

u . :n i 2.msa~L. ;=pe=raties) w v L=o.w-L..o. ~ .

s. 7.st. e.m;_rori

. n -ma, nua. o "W.~es J Th :r a-a w:.w esett

~ ~ ~ -+and.?A. DS -- f funct.i.

.a- o.=aw nstare edif p 4~e._~c'ti=ons.~v372,x w-S

~ w?di=s.~c:u:=s..s_cm'Grthe=rr~in

--- - .ws: a. nd .

' The ' safety function is- performed by the same safety / relief valve with self-actuated integral bellows and pilot valve causing main valve operation. ' Article 9 of the ASME Pressure Vessel Code Section III Nuclear Vessels requires that these bellows-be monitored for failure since this would. defeat the safety function of the safety / relief valve.

Provision also has been made to detect failure of the bellows monitoring system. Testing of this syst,em quarterly provisions assurance.of bellows integrity.

, When the setpoint is being bench checked, it is prudent to disassemble one'of the safety / relief i . valves to examine for crud buildup, bending of certain actuator members or other signs of possible d

'eterioration.

1

~

Low-Low Set LLogic has been provided on three non-Automatic Pressure Relief. System valves.

This. logic is discussed in detail'in the Section 3.2 Bases. This logic,'through pressure sensing

! instrumentation, reduces the opening setpoint and increases.the blowdown range of the three selected valves following a scram to eliminate the discharge line water leg. clearing loads resulting from multiple valve openings.

I. Deleted 1

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3.6/4.6 BASES 151

)

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. . . - . .- _ ~ - ~ . . - - - . - - - .~ . - . - -.. - - - . - - - - -. - .. . .

4 3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Apolicability: Applicability-Applies to the operating status of the primary Applies to the primary and secondary and secondary containment systems. containment integrity.

Obiective: Obiective:

To assure the integrity of the primary and To verify the integrity of the primary and

' secondary containment systems'. secondary containment.

Specification: Specification:

A. Primary Containment. A. Primary Containment .

I

1. Suppression Pool Volume and 1. Suppression Pool Volume and Temperature Temperature When irradiated fuel is in the reactor a. The suppression chamber water vessel and either the reactor water temperature shall be checked once temperature is greater than 212*F or work per day.

is being done which has the potential to b. Whenever there is indication of drain the vessel, the following relief valve operation which adds requirements shall be met, except as heat to the suppression pool, the permitted by Specification 3.5.Ef2G-4: pool temperature shall be ^

~ ~ " ~

continually monitored and also

a. Water temperature during normal observed and logged ever 5 minutes operating shall be 590 F. until the heat addition is

b. Water temperature during test .

terminated.

operation which adds heat to the c. A visual inspection of the suppression pool shall be $100*F and suppression chamber interior shall not be >90*F for more than 24 including water line regions and hours. the interior painted surfaces above

- c. If the suppression chamber water the water line shall be made at temperature is >110"F, the reactor each refueling outage 9

shall be scrammed immediately. Power operation shall not be resumed until the pool temperature is $90 F.

156 3.7/4.7

4 i

Exhibit C Monticello Nuclear Generating Plant l

License Amendment Reauest Dated February 12. 1993 1

Revised Monticello Technical Specification Pages U

i Exhibit C consists of revised Technical Specification pages that incorporate l the proposed changes. The pages included in this exhibit are:

1 k

' 52 1 53 54 55 60d 101 i 107 j'

110 113

• 127 I

151

156 l

r I

e i

i l

_ _ _ _ . _ _ _ _ _. , _~

_ . _ . . __ . . _ _ . ~ _ .. _ . _ _ . . . . .. . . . . _ . . _ . . - _ . _ . _ _ _ . _ _ . . - . _ . . _ _ _ . . . _ _ . . . . _ . _ .. ..

, - i 1

Table 3.2.2 Instrumentation That Initiates Emergency Core Cooling Systems Minimum No. of Oper-Minimum No. of able or Operating Total No. of Instru- Instrument Channels Operable or . l Operating Trip . ment Channels Per Per Trip System Required Trio Settine Systems (3) Trio System (3) Conditions

• I action A. Core Sorav and LPCI i

1. Pump Start-  :

4

a. Low Low Reactor 26'6"s6'10" 2 4(4) 4 A.

Water Level and

b. 1. Reactor Low 2450 psig 2 2(4) 2 A. ,

Pressure Permissive or l

11. Reactor Low- 2011 min 2 1 1 B. i Pressure i  !

Permissive [

Bypass Timer l t

3 c. liigh Drywell 52 psig 2 '4(4) 4 A.

i Pressure (1) i

2. Low Reactor Pressure 2450 psig .2- 2(4) 2 A.

(Valve Permissive)  !

3. Loss of Auxiliary ------ 2 -2(2)~ 2 A.

Power ,

I i

t 2

3.2/4.2 52 1

i

..m_._ .. . .- .- - ._ . . . - . _ _ _ . . . , _ . . . . _ _ . - . _ . _ . . . . _ _ _ _ _ _ . . . . . _ . _ . _ _ . . _ _ . _ . _ _ _ . . . . _ . . . . . . . __

1

- i i

i

- i Table 3.2.2 .

Instrumentation That Initiates Emergency Core' Cooling Systems  !

Minimum No. of Oper- ,

Minimum No of able or Operating +

, Operable or Total No. 'of Instru- Instrument Channels ,

Operating Trip ment Channels Per Per Trip System Required  !

' Function Trio Settina Systems (3) Trin System (3) Conditions

• i B. HPCI System i

, 1. High Drywell. 52 psig. 1- 4 4 A. I Pressure (1)  !

j' '

4 2.

Low-Low Reactor 26'6"s6'10" 1 4 4 A. 'l Water Level C. . Automatic Deores-

surization i

l 1. Low-Low Reactor 26'6"s6'10" 2 2 2 B. l

Water Level  !

and  !

4 4

2. Auto Blowdown s120 seconds 2 1 1 B.

. Timer r 6

and

3. Low Pressure Core $100 psig 2 12(4) 12(4) B. I Cooling Pumps Dis-

~

?  ;

i- . Charge Pressure  ;

Interlock  ;

Y

, 4 [

! :r l t 1 3

i:

i I j- 3.2/4.2 53 .l

.! t

t ..

e i

Table 3.2.2 - Continued Instrumentation That Initiates Emernency Core Coolinn System Min. No. of Oper-Min. No. able or Operating of Operable Total No. of Instru- Instrument Channels or Operating .

ment Channels Per Per Trip System Required Function Triu Settine Trio Systems (3) Trio System (3) Conditions

• D. Diesel Cenerator i

1. Degraded or Loss of Voltage Essential Bus (5)

2. Low Low Reactor 26'6"s6'10" 2 4(4) 4 C.

Water Level

3. Iligh Drywell Press 52 psig 2 4(4) 4 C.

WOTES:

1. High drywell pressure may be bypassed when necessary only by closing the manual containment isolation valves during ,

purging for containment inerting or de-inerting. Verification of the bypass condition shall be noted in the control room log. Also need not be operable when primary containment integrity is not required.

2. One instrument channel is a circuit breaker contact and the other~is an undervoltage relay.

i 3.2/4.2 54 L _

.. . ., _ . _ _ _ _ _ . __ ._..m . _ . _ . - . _ . . _ _ . _ _ _ _ . _ - _ _ . . _ . _ . . _ _ . _ . _

w .l Table 3.2.2 - Continued s

Notesi ,

3. Upon discovery that minimum requirements for the number of operable or operating trip systems, or instrument channels t are not satisfied action shall be initiated to:

(a) Satisfy the requirements by' placing appropriate channels or systems in the tripped condition, or (b) Place the plant under the specified required conditions using normal operating procedures.

~

4. All instrument channels are shared by both trip systems.

5. See table 3.2.6.

O Required conditions when minimum conditions for operation are not satisfied.

A. Comply with Specification 3.5.A.

B. Reactor pressure 5150 psig.

C. Comply with Specification 3.9.B.

j S

i b

1 3.2/4.2 55 i

~ -. . . - . . - . . . . . . _ - . _ . - _ _ . - . . . . - - - . . - - . . . . . - . - , _ . . , - . - ~. - - . - - - - - . . _ . . . . -

6 r

Table 3.2.8 ,

Other Instrumentation ,

Minimum No. of Minimum No. of Oper-Operable or Total No. of Instru- able or Operating Required

' Function- . Trip Setting Operating Trip ment Channels Per Instrument Channels Conditions

• System (1) Trio System Per Trio System (1)

-)

A. RCIC Initiation

1. Iow-Low Reactor Level 26'6"& $6'10" 1 2 2 B above top of active fuel B. HPCI/RCIC Turbine Shutdown

a. liigh Reactor Level 514'6" above 1 2 2 A

- top of active fuel C. HPCI/RCIC Turbine Suction Transfer e a. Condensate-Storage: .22'0" above 1 2 2 C Tank Low Level tank bottom i

i NOTE:

1. Upon discovery that minimum requirements for the number of operable or operating trip systems or -instrument channels are not satisfied, action shall be. initiated to:

a. Satisfy the requirements by: placing the appropriate channels or systems in the tripped condition:

(Turbine /Feedwater Trip only). or-Place the plant under the specified required condition using normal. operating procedures.

!. . b.

c- Required conditions when minimum c,onditions for operation are not satisfied

~

]- A. Reactor in Startup, Refuel, or Shutdown Mode.

B. Comply with Specification 3.5.D.  !

C. Align llPCI and RCIC suction to the suppression pool. Restore channels to operable status within 30 days or-

. place the plant in Required Condition A.

3.2/4.2 60d i

m __._ . ~ . . . . . . . _ _ _ . . _ _ _ . . _ . _ _ . ._ _.~. _ . _ . _ _ _. _ _ _ _ ...m. .. _ . _ _ _ _ __ _ __ _.

~

i 3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT SPRAY / COOLING SYSTEMS 4.5 CORE AND CONTAINMENT SPRAY / COOLING SYSTEMS Applicability: Aeolicability: '

Applies to the operational status of the emergency Applies to the periodic testing of the emergency cooling systems. '!

cooling systems.

Obiective: Obiective:

To insure adequate cooling capability for heat removal To verify the operability of the emergency cooling in the event of a loss of coolant accident or systems, isolation from the normal reactor heat sink.

Specificationi Specification:

A. ECCS Systems A. ECCS Systems 'l

-1. Except as specified in section 3.5.A.3,.both 1. Demonstrate the Core Spray Pumps develop a Core Spray subsystems and the Low Pressure 2,800 gpm flow rate against a system head 3"

Coolant . Injection. (LPCI) Cubsystem (LPCI Mode corresponding to a reactor pressure of 130

of EUEt System) shall be .operchle whenever

~

psi greater than containment pressure, when l _

irradiated fuel'is in the reactoc vessel and tested pursuant to Specification 4.15.B.

the reactor water temperature is greater than 212 F. 2. Demonstrate the LPCI Pumps develop a 3,870

, gpm flow rate against a system head

. 2. Except as specified in section 3.5. A.3, the corresponding to two pumps delivering 7,740 ,

High Pressure Coolant Injection (HPCI) System gpm at a reactor' pressure of 20. psi greater  !

and the Automatic Depressurization System than containment' pressure, when tested  ;

-(ADS).shall be operable whenever the reactor pursuant to Specification.4.15.B.

pressure is greater than 150 psig and irradiated fuel is in the reactor vessel 3. Demonstrate the HPCI Pump develops a 2700 ,

except during reactor vessel hydrostatic or gpm flow rate against a reactor pressure j leakage tests. range of 1120 psig to 150 psig, when tested J

pursuant to Specification 4.15.B.

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l 3.5/4.5' 101.

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3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS F. Recirculation System F. Recirculation System *

l. The reactor may be started and operated, or 1. See Specification 4.6.G operation may continue with.only one recirculation loop in operation provided that: 2. The following baseline noise levels will be obtained prior to operation with only one

a. The following changes to setpoints and recirculation pump in operation at a core .

safety limit settings will be made within thermal power greater than that specified 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after. initiating operation with in Figure 3.5.1 or with a core flow greater only one recirculation loop in operation. than 45% provided that baseline values have not been established since the last core 1.'The Operating Limit MCPR (MCPR) will be refueling. Baseline values will be taken ,

changed per Specification 3.ll.C.

with only one recirculation pump running.

2.'The Maximum Average Planar Linear Heat a. Establish a baseline core plate AP noi.se Generation Rate (MAPUICR) will be level.

changed as noted in Table 1 of the Core .I Operating Limits Report. b. Establish a baseline APRM and LPRM neutron flux noise level.

3. The APRM Neutron' Flux Scram and APRM Rod Block setpoints will be changed as noted 3. With only one recirculation loop in in Specification 2.3.A and Table 3.2.3. operation at a core thermal power greater than that specified in Figure 3.5.1 or with

b. Total core flow will be. maintained greater a core flow greater than 45%, determine the than 39% when core thermal po'wer is above following noise levels at least once per 8 the limit specified in Figure 3.5.1. hour period and within 30 minutes after a core thermal. power increase of greater than 5% of rated thermal power.

a. Core plate AP noise levels.

4 t

b. APRM and LPRM neutron noise levels.

3.5/4.5 107 i

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Bases 3.5/4.5 A. ECCS Systems The core spray system is provided to assure that the core is adequately cooled following a loss-of-coolant accident and, together with the LPCI mode of the RHR system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the automatic depressurization system (ADS). ,

The Core -Spray System is a primary source of emergency core cooling after the reactor vessel is i depressurized and a source for flooding of the core in case of accidental draining. The Core Spray pump is designed to deliver greater than or equal to 3020 gpm (the SAFER /GESTR-LOCA safety analysis assumed a Core Spray Pump flow of 2,800 gpm, or 2,700 gpm flow into the core + 100 gpm to account for ECCS bypass leakage) against a system head corresponding to a reactor pressure of 130 psi greater than containment pressure.

The surveillance requirements provide adequate assurance that the Core Spray System will be operable when '

required. Although all active components are testable and full flow can be demonstrated by recirculation '

.through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment. i

.i The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is ,

adequately cooled following a loss-of-coolant accident. Four pumps are available to provide adequate core "

flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS. LPCI Loop Selection Logic determines which r Recirculation loop the four RHR pumps will pump into. Each FGUt pump was designed to deliver greater than or equal to 4000 gpm (the: safety analysis assumed two pumps delivering 7,740 gpm) against a system head r i corresponding to a reactor: pressure of 20 psi greater than containment pressure.

The allowed out-of-service conditions (Section 3.5. A.3) are determined from ECCS analysis cases analyzed. l Only one of these conditions-is permitted to exist. If more than one condition exists, an orderly shutdown shall be initiated. A LPCI injection path consists of the two rotor operated injection valves on that path.

Th'e surveillance requirements provide adequate assurance that the LPCI system will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation ,

through a test loop during reactor operation, a complete functional test requires reactor shutdown. The  !

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3.5/4.5 110 1

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Bases 3.5/4.5 Continued:

The RHR service water system provides cooling for the RHR heat exchangers and can thus maintain the suppression pool water within limits. With the flow specified, the pool temperature limits are maintained  ;

as specified in Specification 3.7.A.l.

i D. RCIC The RCIC system is provided to' supply continuous makeup water to the reactor core when the reactor isolated from the turbine and when the feedwater' system is not available. The pumping capacity of the RCIC system is  ;

sufficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts.

The system may also be manually initiated at any time.

The HPCI system provides an alternate method of supplying makeup water to the reactor should tho normal ,

feedwater become unavailable. Therefore, the specification calls for an operability check of the HPCI system should the RCIC system be found to be inoperable.

The surveillance requirements provide adequate assurance that the RCIC system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge piping is maintained full to prevent water hammer

' damage and to provide cooling at the earliest moment.

E. Cold Shutdown and Refueling Requirements The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times. It is during refueling outages that major maintenance is performed and during such time that all' core and containment spray / cooling subsystems may be out of service. This specification allows all core and containment spray / cooling subsystems to be inoperable provided no work is being done which has the potential for draining f.he reactor vessel. Thus events requiring core cooling are precluded.

. Specification 3.5.E.2 recognizes'that concurrent with control rod drive maintenance during the j

' refueling outage, it may by,necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7.A.l. In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.

i 3.5/4.5 Bases 113 l .

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e 3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS E. . Safety / Relief Valves E. Safety / Relief Valves

1. During power operating conditions and whenever 1. a. A minimum of seven safety / relief valves reactor coolant pressure is greater _than 110 shall be bench checked or replaced with psig and temperature is greater than 345'F the a bench checked valve each refueling safety valve function (self actuation) of outage. The nominal self-actuation seven safety / relief valves shall be operable setpoints are specified in Section (note: Low-Low Set and ADS requirements are 2.4.B.

located in Specifications 3.2.H and 3.5.A, respectively). b. At least two of the safety / relief valves shall be disassembled and  ;

2. If Specification 3.6.E.1 is not met, initiate inspected each refueling outage.

an orderly shutdown and have reaccor coolant ,

pressure and temperature reduced to 110 psig c. The integrity of the safety / relief t or less and 345*F or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. valve bellows shall be continuously monitored.

d. The operability of the bellows monitoring system shall be demonstrated at least once every three months.

2. Iow-Low Set Logic surveillance shall be performed in accordance with Table 4.2..l.

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Bases Continued 3.6 and 4.6:

. The safety / relief valves have two functions: 1) over-pressure relief (self-actuated by high pressure), and 2) Depressurization/ Pressure Control (using air actuators to open the valves via ADS, Low-Low Set system, or manual operation). The Low-Low Set and ADS functions are discussed further in Sections.3.2 and 3.5.

The safety function is performed by the same safetyfrelief valve with self-actuated integral bellows and pilot valve causing main valve operation. Article 9 of the ASME Pressure Vessel Code Section III Nuclear Vessels requires that these bellows be monitored for failure since this would defeat the safety function of the safety / relief valve. ,

Provision also has been made to detect failure of the bellows monitoring system. Testing of this system quarterly provisions assurance of bellows integrity.

When the'setpoint is being bench checked, it is prudent to disassemble one of the safetyfrelief valves to examine for crud buildup, bending of certain actuator members or other signs of possible deterioration.

Low-Low' Set Logic has been pr$vided on three non-Automatic Pressure Relief System valves.

This logic is discussed in detail in the Section 3.2 Bases. This logic, through pressure sensing instrumentation, reduces the opening setpoint and increases the blowdown range of the three selected valves following a scram to eliminata the discharge line water leg clearing loads resulting from multiple valve openings.

I. Deleted-1 1

3 3.6/4.6 BASES 151 4

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s 3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIldCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Applicability: Applicability:

Applies to the operating status of the primary Applies to the primary and secondary and secondary containment systems. containment integrity.

Obiective: Obiective:

To assure the integrity of 'he primary and To verify the integrity of the primary and secondary containment systems. secondary containment.

Specification: Specification:

A. Primary Containment. A. Primary Containment

1. Suppression Pool Volume and 1. Suppression Pool Volume and Temperature Temperature When irradiated fuel is in the reactor a. The suppression chamber water vessel and either the reactor water temperature shall be checked once temperature is greater than 212*F or work per day.

is being done which has the potential to b. Whenever there is indication of drain the vessel, the following relief valve operation which adds ,

requirements shall be met, except as heat to the suppression pool, the permitted by Specification 3.5.E.2: pool temperature shall be continually monitored and also

a. Water temperature during normal observed and logged ever 5 minutes operating shall be 590*F. until the heat addition is

b. Water temperature during test terminated.

operation which adds heat to the c. A visual inspection of the suppression pool shall be $100*F and suppression chamber interior shall not be >90 F for more than 24 including water line regions and hours. the interior painted surfaces above

c. If the suppression chamber water the water line shall be made at temperature is >110 F, the reactor each refueling outage shall be scrammed immediately. Power operation shall not be resumed until the pool temperature is $90 F.

156 3.7/4.7

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