Text

Proposed Tech Specs Pages,Incorporating Ultimate Heat Sink Temperature of 75 F Into TS 3/4.5.B & Bases,As Required by Amend 173
	ML20203D498
Person / Time
Site:	Pilgrim
Issue date:	02/20/1998
From:	; BOSTON EDISON CO.
To:	;
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	NUDOCS 9802260079
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ATTACHMENT A Markup of Current Technical Specifications-i d

9002260079 980220 PDR ADOCK 05000293 P PDR

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lk1MITING CONDITION FOR O?ERATION SURVEILLANCE REQUIREMENT l3.5 CORE AND CONTAlfMENT COOLING 4.5 CQRE AND CONTAf fMEN r CMLIN3 ;

SYSTEMS SYSTD45 I

? !

I Aeolicability Aeolicability Applies to the operational status Applies to the Surveillance f

of the core and suppression pool Requirements of the core and cooling systems, suppression pool cooling systems l which are required when the ;

corresponding Limiting condition j for operation is in effect.

l

/h[1\3 Obiective Obiective c

To assure the operability of the To verify the operability of the l core and suppression pool cooling core end su;pression pool cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capability is an essential response to station an essential response to station abnormalities, abnormalities.

Soecification Specification A. Core Sorav and LPCI Systems A. Core Sorav and LPCI Systems

1. Both core spray systems shall 1. Core Spray System Testing, be operable whenever irradiated fuel is in the vessel and prior 112m Frecuency to reactor startup from a Cold Condition, except as 'pecified a. Simulated Once/

in 3.5,A.2 below. Automatic Operating Actuation test. Cycle

b. Pump When tested Operability as specified in 3.13 verify that each core spray pump j delivers ;

at least l

3300 GPM i I

against a j system head; correspond-l ing to a reactor vessel pressure of 104 psig l

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Revi4+ ion-aAF7 !

Amendment No, i?. '?, 'l;, ' ' . . 1F. '; 3/4.5-1 L_. - - - - - - . _ . . . . . . _ - - _ _ . .- .-. - - _ - . . . . _ - - .

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klMLTIHO CONDITION FOR OPERATION SURVEILLANfg REOUIEEMENT 3.5 COPE AND CONTAINMENT CTLIll'3 4.5 CORE AND CONTAltlMENT CMI,HQ SYSTEMS (Cont) SYSTEd3 (Cont)

A. Core Sorav and IPCI Systems A. Core Sorav and LPCI Systems (Cont) (Cont)

2. From and ofter the date that c. Motor Operated As specified one of the core spray systems valve in 3.13 A1 is made or found to be Ope.sbility e

inoperable for any reason, continued reactor operation is d. Core Spray Header permis-ible during the Ap Instrumentatic n succeeding seven days, provided that during such seven days all Check once:'iay active components of the other core Fpray system and active Calibrate Once/3 components of tha LPCI system months and the diesel ger.erators are operable. Test Step Once/3 months

3. The LPCI system shall bt?

operable whenever irradiated 2. This sect ion intentionally lef t fuel is in the reactor vessel, blank and prior to reactor startup from a Cold Condition, except 3. LPCI system testing shall be as as specified in 3.5.A.4 and follows:

3.5.F.5,

a. Simulated Once/

4. From and after the date that Automatic Operating the LPCI system is made or Actuation Cycle found to be inoperable for any Test reason, continued reactor operation is permissible only b. Pump When tested during the succeeding seven Operability as specified days unless it is sooner made in 3.13 operable, provided that during verify that such seven days the lmeme nt each LPCI mt% ) c,ctm ;d i 4 e pump At W" r n;*+ ed-betive delivers components of both core spray 4800 GPM systemsr and the diesel at l generators required for a head operation of such components if across the no external source of power pump of were available, shall be at least operable. 380 feet

5. If the requirements of 3.5.A c. Motor Operated As specified cannot be met, an orderly valve in 3.13 shutdown of the reactor shall operability be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Revision 17' Amendment No. 'O, ', ~2, ' '. 1 ?, 1". 144 3/4.5-2

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r o-LIMITING CONDITION FOR OPERATION AURVIILLANCE REQUIREMENT 3.5 CQIB AND COffTAltMENT COOL 1NG 4.5 . CORE A!!D_CONTAf tMEtJT COOLIt{g

' SYSTEMS (Cont) SYSTEMS (Cont)

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Amendment No, 30, 12, : : , ;1;, ;35, 110 3/4.5 l l

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GaunD LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT SYSTEMS COOLING SYSTEMS

^ - -

B.1 Re.gidual Heat Removal (RHR) _ B.1 Residual Heat Removal (RHR)

Suporession Pool Coolino Suppression Pool Cooling Specification:

Two RHR suppression pool cooling 1. Verify each RHR suppression .

subsystems shall be OPERABLE. pool cooling subsystem manual, power operated, and automatic valve in the flow path that is Applicability: not locked, sealed, or otherwise 4 secured in position is in the Whenever irradiated fuel is in the reactor correct position or can be aligned vessel, reactor coolant temperature is to the correct position every

> 212' F, and prior to startup from a cold 31 days.

1 . condition.

Actions: 2. Verify each RHR pump develops a flow rate ;t 5100 gpm through A. ')r.e RHR suppression pool cooling the associated heat exchanger subsystem inoperable, while operating in the suppression pool cooling mode

1. Restore the RHR as specified in Specification

, suppression pool cooling 3/4.13.

subsystem to OPERABLE status within 7 days.

B. Required Action and associated Compietion Time not met, QB Two RHR suppression pool cooling subsystems inoperable,

1. Be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No. 3/4.5-3

Cia =O LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS - _

~ -

B.2 Residual Heat Removal (RHR) B.2 Residual Heat Removal (RHR) l Containment Soray Containment Soray Specification:

Two RHR containment spray subsystems 1. Verify each RHR containment shall be OPERABLE. spray subsystem manual, power operated, and automatic valve yl Aoolicability: in the flow path that is not locked, sealed, or otherwise Whenever irradiated fuelis in the reactor secured in position is in the vessel, reactor coolant temperature is correct position or can be

> 212 F, and prior to startup from a cold aligned to the correct position M 4 condition-everv. 31 dav.s-Mtions: ) 2. Air test drywell and suppression pool (torus) headers and A, One RHR containment spray subsysteni nozzles once per 5 years.

inoperable, I 1. Restore the containment b spray subsystem to OPERABLE status within 7 days, B- Required Action and associated Completion Time not met, m /

Two RHR containment spray ,

subsystems inoperable,

1. Be in Cold Shutdown within 24 ho Amendment No. 3/4.5-4

LIMITING CONDITIONS FOR OPERATION b

SURVEILLANCE REQUIREMENTS A. 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT SYSTEMS COOLING SYSTEMS

}6 Reactor Buildina Closed Coolina lA (RBCCW) System B.3 Reactor Water Buildina Closed Coolin)

Water (RBCCW) System Specification:

o RBCCW subsystems shall be h1. -

-NOTE Isolation of flow to individual OPERABLE.

\

components does not render the A_policability: RBC CW subsystem inoperable.

,g r'Whenever irradiated fuel is in the reacto

( .

J enfy V each RBCCWmanual,\

vessel, reactor coolant temperature is

/ power operated, and automatic

> 212 F, and prior to startup from a cold valve in the flow path, that is not condition.

locked, sealed, or otherwise Ma As secured in position, is in the

(_

,1 Actions] - correct position or can be aligned to the correct position A. One required RBCCW pump every 31 days. ,

inoperable, *

y 1. Restore the required RBCCW pump to OPERABLE status /

within 7 days. j s

One RBCCW subsystem inoperable [1 1, Restore RBCCW subsystem A,- to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

C. Required Action and associated Completion Time not met, DB Two RBCCW subsystems OA,1 inoperable, - \

1. Be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1 I

l Amendment No. 3/4.5-5

LIMITING CONDITIONS FOR OPERATION CmD SURVEILLANCE REQUIREMENTS 3.5 _QORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT SYSTEMS COOLING SYSTEMS B.4 Salt Service Wate SSW)1 System and B.4 Salt Service Water (SSW)1 System !

Ultimate Heat Sink (UHS) and Ultimate Heat Sink fyliS)

Specification; f -

Two SSW subsystems shall be 1. Venfy the water levelin each 4* OPERABLE. SSW pumpwell of the intake M4 structure)is 213 ft. 9 in.below Applicability; mean sea level every 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,s/

IWheney irra ed fuel is in the reactor 2. Venfy the average sea j vessel, reactor coolant temperature is temperature is s 75'F every 24j

b. > 212' F, and prior to startup from a cold hours.j condition. f A8- Actions: 3. -- -NOTE

, Isolation of flow to individual A One SSW subsystem inoperable, Ae , components does not render s the SSW subsystem inoperable.

J

1. Restore the SSW subsystem _

m - -

to OPERABLE status within j Venfy each SSW subsystem 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . J manual, power operated, and a

~

automatic valve in the flow B. Required Action and associated

paths servicing safety- related (h Completion Time not met, I

systems or ;omponents, that is not locked, sealed, or otherwise gg ured in sition n the Two SSW subsystems inoperable, 4 ,

I UHS inoperable, O ,

('1. Be in Cold Shutdown within 4. Verify each SSW subsystem

\ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . actuates on an actual or sin.ulated initiation signal every 2 years p

@J Amendment No. 3/4.5-6

LIMITING CONRIIION FOR OPERATION AURVEILLANCE REQUIREMElII l 3.5 CORE AND CONTAINMENT COOLIN'] 4.5 COPE AND CONTAINMENLCnOLING SYSTEMS (Cont) SYSTEMS (Cont)

C. HPCI System C. HPCI System

1. The HPCI system shall be 1. HPCI system testing shall be operable whenever there is performed as follows:

irradiated fuel in the reactor vessel, reactor pressure is

a. Simulated Once/

greater than 150 psig, and Automatic operating reactor coolant temperature is Actuation cycle greater than 365'F except as Test specified in 3.5.C.2 below,

b. Pump Oper- When tested 2 From and after the date that ability as the HPCI system is made or specified A, found to be inoperable for any in 3.13 reason, continued reactor verify operation is permissible only that the during the succeeding 14 days HPCI pump unless such system is sooner delivers at made operable, providing that least 4250 during such 14 days all active GPM for a components of the ADS system, system head the RCIC system, the LPCI correspond-system and both core spray ing to a systems are operable, reactor pressure of

3. If the requirements of 3.5.C psig, cannot be met, an orderly shutdown shall be initiated and

c. Motor Operated As the reactor pressure shall be reduced to or below 150 psig Valve Oper- specified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , ability in 3.13

d. Flow Rate at Once/

150 psig operating cycle verify that the HPCI pump delivers at least 4250 GPM for a system head correspond-ing to a reactor pressure of 150 psig.

The HPCI pump shall deliver at least 4250 gpm for a system head corresponding to a reactor pressure of 1000 to 150 psig.

Rev . .l _

'^^

Amendment No.-3R, l', 44, , l':, '"

,, 144, i ' ,

3/4.5 4 7

. 1 LJji1 TING CONDITION FOR O.EKEA?LQ){ SURVEIkl.ANCE REQUIREMENT l

l 3.5 CORE AND COf1TAlfiMENT COOLif1'] 4.5 CORE Af1D COf1TAf fiMEf1T COOLIfiG SYSTEMS (Cont) SYSTEMS (Cont) i D. Reactor Core Isolation Cooling D. Reactor Core Isolation Coolina (RCIC) System 1ECJC) System

1. The RCIC system shall be 1. RCIC system testing shall be operable whenever there is performed as follows:

irradiated fuel in the reactor vessel, reactor pressure is a. Simulated Once/operatirg greater than 150 psig, and Automatic cycle reactor coolant temperature is Actuation greater than 365'F; except es Test specified in 3.5.D.2 below,

b. Pump When tested

2. From and after the date that Operability as the RCICS is made or found to specified be inoperable for any reason, in 3.13 continued reactor power veriff that operation is permissible only the RCIC during the succeeding 14 days pump provided that during such 14 delivers days the HPCIS is operable, at least 400 GPM

3. If the requirements of 3.5.D at a cannot be met, an orderly system head At shutdown shall be initiated and correspond-the reactor pressure shall be ing to a reduced to or below 150 psig reactor within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , pressure of 1000 poig.

c. Motor As specified Operated in 3.13 Valve Operability

d. Flow Rate at once/ operating 150 psig cycle verify that the RCIC pump delivers at least 400 GPM at a m/ stem head correspond-ing to a reactor pressure of 350 psig.

The RCIC pump shall deliver ar least 400 gpm for a system head corresponding to a reactor pressure of 1000 to 150 psig.

4k v+*hm--H4 Amendment No.-42, '^9, 'l*, 135, ; O;, 4 % 3/

LI]ilIJHG CONDITION FOR OPERATION SURVELLLANCE REQUIREMElfI 1.5 CORE AND CONTAlfNENT COOLIN'] 4.5 CORE AND CONTATfMENT ContING SYSTEMS (Cont) SYSTEMS (Con')

R. Automatic Der.ressurization System E. Automatic Derressurization System (ADS) (ADS)

1. The Automatic Depressurizatior. 1. During each operating cycle the System shall be operable following tests shall be

4, whenever there is irradiated performed on the ADS:

fuel in the reactor vessel and the reactor pressure is greater a. A simulated automatic than 104 psig and prior to a actuation test shall be startup from a Cold Condition, performed prior to startup except as specified in 3.5.E.2 after each refueling outage.

below. The ADS manuel inhibit switch will be included in

2. From and after the date that this test, one valve in the Automatic Depressurization System is made b. With the reactor at or found to be inoperable for pressure, each relief valve any reason, continued reactor shall be manually opened operation is permissible only until a corresponding change during the succeeding 14 days in reactor pressure or main unless such valve is sooner turbine bypass valve made operable, provided that positions indicate that during such 14 days the HPCI t;eam is flowing from the system is crierable, valve.

3. If the requirements of 3.5 E cannot be met, an orderly shucdown shall be initiated and the reactor pressure shall be reduced ~o

at least 104 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

%+kMtm-+-M Amendment No. '2, 42, :7 C, ' ^ r: , lla, ' 1, :=' .5-9 3/4_J]

LIMITING _QJ;WDITION FOR_QEgMATlOJ SURVEILLANCE Bg_QUIREMENT 3.$ GQEE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLINQ SYSTEMQ (Cont) SYSTEMS (Cont)

! F. F, MLn] mum Low Pressure Coolina and Diesel &nimum Low Pressure Coolina and Diesel

Generator Availabildy generator Availabildv i 1. During any period when one diesel 1. When it is determined that one diesel l generator is inoperable, continued reactor generator is inoperable, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , operation is permissible only during the determine that the operable diesel generator

succeeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months unless such diesel is not inoperable due to a common cause generator is sooner made operable, provided failure, that all of the low pressure core and containment cooling systems and the 91 l remaining diesel generator shall be operable. If this requirement cannot be inet, perform surveillance 4.9.A.I.a for the an orderly shutdown shall be initiated and operable diesel generator, the reactor shall be placed in the Cold i

Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , gnd ,

2. Any combination of inoperable components within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months In the core and containment cooling systems thereafter, venfy correct breaker alignment shall not defeat the capability of the and indicated power availabildy for each fornaining operable components to fulfill the offslie circult.

cooling functions.

! 3. When irradiated fuelis in the reactor vessel and the reactor is in the Cold Shutdown At condition, both core spray systems, the i LPCI and containment cooling systems may be inoperable, provided no work is being done which has the potential for draining the reactor vessel.

4. During a refueling outage, for a period of 30 days, refueling operation may continue provided that one core spray system or the

L,0Cl system is operable or Specification 3.5.F.5 is met, 4

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Rwesio+403 -

Amendment No.16ctM470 '

3/4.5{7[

e*-ee= -----e,-tTs yvt* v e-- 4 w t+ v v w- i-vtww==wr*y't---mi-e-'-***y 7 a= = --e y ev-- ev W & ----*s --

-***<w'eev--e-ce- w =+-i-1***-e* +c-*-e-* ------t- --

-+e--et=~'--N*-'W- -**iwwe,1r--- -,4,

e LIMITING CONDITION FOR OFIFATION KERYIMcK_REQUIREMitif 3.5 COPE Af1D col {IAltfMEffT COOL 1til 4.$ CORF Af1D CO!JTAlfMEllT COnLI!JG SYSTEMS (Con:) SYSTEMS (Cont)

F, Minimum Low Pressure Coolin, and Diesel Generatcr Availability (Cont)

5. When irradiated fuel 16 in the reactor vessel and the reactor is in the Refueling Condition with the torus drained, a single control rod drive mechanism may be removed, 11 both of the following conditions are satisfied a) No work on the reactor vessel, in addition to CRD A-l removal, will be performed which has the potential for exceeding the maximum leak rate from a single control blade seal if it became unseated, b) 1) the core spray systems are operable and aligned with a suction path f rom the condensate storage tanks.

11) the condensate storage tanks shall contain at least 200,000 gallons of usable water and the refueling cavity and dryer / separator H. Maintenance of Pilled Discharae pool shall be flooded to a least elevation 114'-0* '

The following surveillance G. Deleted requirements assure that the discharge piping of the core spray H. Maintennnce of Filled Discharoe systems, LPCI system, HPCI and RCIC are filled:

Whenever core spray systems, LPCI system, HPCI or RCIC are required 1. Every month the LPCI system and to be operable, the discharge core spray system discharge piping from the pump discharge of pipino shall be vented from the these systems to the last block high point and water flow valve shall be filled. observed.

11 Rev4t+hwt 447 Amendment No.-497-+45r-466 3/4.5j{

0 kIKIIIMLCONDin0H_E91LOIIMn0H avaynLLantLMQuntMKHI 3.5 COPE /d1D COfJTAIfff4EffT CnOLIIJ'1 4.5 COPE /dlD COfJTAIfI!EffT COntIfJ'1 SYSTDMS (Cont) SYSTilf49 (Cont)

H. 14aint enance of Filled Dischsr'ie Eirg (cont)

2. Following any period where the LPCI system or core spray systems have not been required to be operable, the discharge piping of the inoperable system ,

shall be vented from the high At point prior to the return of the system to service.

3. Whenever the HPCI or RCIC system is lined up to take suction from the torus, the discharge piping of the HPCI j and RCIC shall be vented from i the high point of the system and water flow observed on a monthly basis. I i

l .

w w w as Amendment No.-Mr-HWW 3/4.5]

l ATTACHMENT B Discussion of Changes

j . DISCUSSION OF CHANGES CURRENT TS (CTS) 3.5 CORE AND CONTAINMENT COOLING SYSTEMS ADMINISTRATIVE CHANGES !

Ai These proposed changes include reformatting (font and style changes) and j repagination only. No technical changes (either actual or interr:7tational) to the TS were made.

A This change proposes to delete ...

containment cooling system (including 2 LPCI pumps) and"... from Specification 3.5.A.4. The originalintent of specifying containment cooling requirements within the LPCI specification was to ensure that stifficient pumps remained available to provide the suppression pool cooling i

safety function of the RHR system. The proposed, new specifications 3/4.6.B.1

RHR Suppression Pool Cooling System" and 3/4.5.B.2 *RHR Containment Spray System will more clearly define the requirements for the containment '

' cooling functions of the RHR system. Therefore, this change is considered '

idministrative.

A This change proposes to replace existing Specification 3.5.B

Containment Cooling System" with new Specifications 3/4.5.B.3

Reactor Building Closed j Cooling Water SysterT and 3/4.5.B.4 " Salt Service Water System".

All editorial rewording (either adding or deleting), reformatting and renumbering is proposed to restructure the section to account for the additional Specifications, ACTIONS, and Surveillances. These proposed changes are intended to result in Technical Specifications (TS) that are more readable and, therefore, understandable by plant operators as well as other users. The

editorial rewording, reformatting and renumbering contains no technical changes (either actual or interpretational) to the TS except those that are identified and justified elsewhere.

Additional changes (Administrative, More Restrictive, Relocations, and Less Restrictive) that require further justification are identified in INSERT 1.

A4 This change proposes to renumber SR 4.5.8.1.c to 4.5.B.2.2 to coincide with the proposed new Specification for Containment Spray. No technical changes (either actual or interpretational) to the requirements were made.

As This change proposes to reword the Specification (SPECIFICATION) of 3.5.B, "both containment cooling system loops shall be operable", to specifically address new Specifications 3/4.5.B.3

Reactor Building Closed Cooling Water System" and 3/4.5.B.4 " Salt Service Water System". These proposed changes '

will result in no technical change (either actual or interpretational) to the Technical Specifications. ,

1 i

. DISCUSSION OF CHANGES CURRENT TS (CTS) 3.5 CORE AND CONTAINMENT COOLING SYSTEMS ADMINISTRATIVE CHARQjiS (continued)

As This change proposes to reformat the applicability of Specification 3.5.B.

"whenever irradiated fuel is in the reactor vessel and reactor coolant temperature is greater than 212' F, and prior to Startup from a Cold Condition

for new Specifications,3/4.5.B.3 ' Reactor Building Closed Cooling Water System", and 3/4.5.B.4

Salt Service Water System". These proposed changes

, Will result in no technical change (either actual or interpretational) to the Technical Specifications.

At These changes propose to reformat the actions required for one inoperable subsystem of the reactor building closed cooling water or salt service water systems. The conditions, required actions and completion times to restore the inoperable subsystem remain the same as in current Technical Specifications.

No technical changes (either actual or interpretational) to the requirements were made.

As This proposed change adds a Note to SR 4.5.B.3.1 and SR 4.5.B.4.3 to clarify that an RBCCWiSSW subsystem need not be considered inoperable when flow to individual components serviced by the RBCCW/SSW system is isolated. As such, when all necessary pumps, valves, heat exchangers, and piping necessary to support the safety functions, the RBCCW/SSW subsystems are operable. This Note is consistent with what constitutes an OPERABLE system as described in the Bases, therefore, this is considered an administrative change.

TECHNICAL CHANGES - MORE RESTRICTIVE Mi TS 3.5.B.1 is a new requirement and requires two RHR suppression pool cooling subsystems to be operable. While it appears, based solely on the title, that current TS (CTS) 3.5.B requires containment cooling, no specific suppression pool cooling requirements, ACTIONS or surveillance requirements (SRs) currently exist. The existing SRs ensure operability of the RBCCW and SSW pumps and valves and drywell and suppression pool spray headers. The RHR suppression pool cooling system is being added since it is a function assumed in transient and accident analyses. Appropriate ACTIONS and SRs are also included. The 7 day completion time was chosen instead of 3 days because (i) it is consistent with other allowable outage times (AOTs) for systems required to mitigate a loss of coolant accident (LOCA) (i.e., core spray and low pressure coolant injection systems), the redundant RHR suppression pool cooling / spray capabilities afforded by the OPERABLE subsystem, and the low probability of a DBA or transient occurring during this period that would require this function, (ii) it is also consistent with CTS 3.5 A.4. which specifies a 7 day AOT for the LPCI system provided the containment cooling system with 2 LPCI l pumps is operable, and (iii) the same pumps are used for both LPCI and containment cooling modes of operation.

l 2

. DISCUSSION OF CHANGES CURRENT TS (CTS) 3.5 CORE AND CONTAINMENT COOLING SYSTEMS TECHNICAL CHANGES MORE RESTRICTIVE (continued)

Mr TS 3.5.B.2 is a new requirement and requires two RHR containment sprhy subsystems to be operable. While it appears, based solely on the Title, that CTS 3.5.B requires containment cooling, no specific suppression pool spray requirements, or ACTIONS currently exist. The existing SRs ensure operability of the RBCCW and SSW pumps and valves and drywell and suppression pool spray headers. The RHR containment spray system is being added since it is a function assumed in transient and accident analyses. Appropriate ACTIONS and SRs are also included. The 7 day completion time was chosen because it is consistent with other AOTs for systems required to mitigate a loss of coolant accident (LOCA) (i.e., core spray and low pressure coolant injection systems),

the redundant RHR suppression pool cooling / spray capabilities afforded by the OPERABLE subsystem, and the low probability of a DBA or transient occurring during this period that would require this function.

M2 These changes propose to add a new Surveillance (4.5.B.3.1, and 4.5.B.4.3) to periodically verify the correct valve position is maintained to ensure the subsystems remain capable of providing the overall cooling requirement. The addition of requirements is a more restrictive change.

M4 This change proposes to add a Surveillance Requirement (4.5.8.4.1) to verify the water level in each SSW pump well. This change adds additional requirements to Technical Specifications which constitute a more restrictive change.

Ms This change proposes to add a Surveillance Requi.ement (4.5.B.4.2) to verify average sea water temperature to assure operab!lity of the Ultimate Heat Sink.

This change adds additional requirements to Technical Specifications wt ich constitute a more restrictive change.

Me This change proposes to add a Surveillance Requirement (4.5.B.4.4) to verify each SSW system actuates on an actual or simulated initiation signal. This change adds additional requirements to Technical Specifications which constitute a more restrictive change.

TECHNICAL CHANGES - RELOCATIONS Ri This change proposes to relocate the pump and motor operated valve testing requirements (4.5.B.1 a and 4.5.B.1.b) to the PNPS IST program (Specification 3/4.13) and the procedures implementing the IST program. Any changes to this testing provision will be adequately controlled by 10 CFR 50.55a and 10 CFR 50.59.

3

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. DISCUSSION OF CHANGES !

i CURRENT TS (CTS) 3.5 CORE AND CONTAINMENT COOLING SYSTEMS ;

IECBNICAL CHANGES LESS RESTRICTIVE .

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i Li This change proposes a new ACTION and Completion Time for one required 2

RBCCW pump inoperable. With one required RBCCW pump inoperable, the remaining pump in the affected loop is sufficient to handle the normal operation l heat loads, and the remaining OPERABLE loop is sufficient to meet the ;.

3

^

requirements to support SPECIFICATIONS 3/4.5 B.1, "RHR Suppression Pool Cooling System" and 3/4.6.B.2 *RHR Containment Spray System". The 7 day 1 Completion Time is consistent with the Completion Times for one Inoperable i loop of suppression pool cooling system or containment spray subsystem.

l Seven days is considered adequate for this reason and the low probability of an [

^

]- event occurring during this period that would require RBCCW to support the containment and core cooling functions.

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4 ATTACHMENT C i

4 j Determination of No Significant Hazard Consideration j

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e DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION CTS 3.5 CORE AND CONTAINMENT COOLING SYSTEMS The Code of Fodoral Regulations (10 CFR 50.91) requires licensees requesting an amendment to provido an analysis, using the standards in 10 CFR 50.92, that determinos whether a significant hazards considoration exists. The following analysis is provided in accordance with 10 CFR 50.91 and 10 CFR 50.92 for the proposed amendment.

ADMINIS]MTlYlLCilANGliS (Ai, Ar A3, A4, As, Ac, Ar and As Labolod Discussion of Changes, CTS 3.5) 1 hose proposed changes involvn rnformatting renumberin0, and roword;ng of the Technical Specifications (TS). Thoso changos, since they do not involve technical changes to the TSs, are administrativo. All of the administrative changes contained in the Discussion of Changes for this chaptor are addressed the following ovalation.

BECo has ovaluated the proposed changos and has determined that they involvo no significant hazards consideration. This dolormination has boon performed in accordance with the critoria set forth in 10 CFR 50.92. The following evaluation is provided.

1. Doos the chango involvo a significant increase in the probability or consequences of an accident previously evaluatod?

Operation of PNPS in accordance with the proposed chango will not involvo a significant increase in the probability or consequences of an accident previously evaluated because of the following:

Thoso proposed changes (editorial rewording, reformatting, ropagination, and renumbering) are made to rostructuro the section, accounting for the now specifications replacing Specification 3/4.5.B. These proposed administrativo changos do not alter any existing requirements.

2. Doos the chango create the possibility of a now or difforent kind of accident from any accident previously evaluated?

Operation of PNPS h accordance with the proposed chango will not croato the possibility of a now or different kind of accident from any accident previously evaluated because of the following:

The proposed changes do not involvo a physical alteration of the plant (no now or different type of equipment will be installed) or changos in methods governing plant operation. The proposed changes will not impose any now or different requirements or climinato any existing requirements.

3. Does this change involve a significant reduction in a margin of safety?

Operation of PNPS in accordance with the proposed change will not involvo a significant reduction in a margin of safety because of tho following:

The changes are administrativo in nature and do not involvo any technical changos.

Sinco no technical changos (either actual or interpretational) were mado, there is no impact on any safety analysis margin of safety, 1

. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION CTS 3.5 CORE AND CONTAINMENT COOLING SYSTEMS IECHNICAL CHANGES MORE RESTRICTlyfi (Mi, M,, M3, M , Me, and Me Labeled Discussion of Changes, CTS 3.5)

Those proposed changos incorporato more rostrictivo changes into the current Technical Specifications by either making current requirements more stringent or adding now requirements which currently do not exist. All of the more rostrictive changes contained in the Discussion of Changes for this chapter are addressed by the following ovalt.ation.

DECO has eva!uated the proposed Technical Specification changes and has dolormined that they involvo no significant hazards consideration. This dolormination has boon performed in accordance with the critoria sot forth in 10 CFR 50.02. The following evaluation is provided.

1. Does the change involvo a significant increase in the probability or consequences of an accident previously evaluatod?

Operation of PNPS in accordance with the proposed changos will not involvo a significant increase in the probability or consequencos of an accident previously evaluated becauso of the following:

The proposed changos provido more stringent requ;romonts than previously existed in the Technical Specifications. The more stringont requiroments provide greator emurance that the affected systems will remain capable of providing the safety functions assumod in design basis accidents and transients. If anything, the now requirements may decrease the probability or consequences of an analyzed event.

The change will not alter assumptions relativo to mitigation of an accident or transient ovent. The more restrictive requirements will not alter the operation of process variablos, structures, systems, or components as described in the safety analyses.

2. Does the chango croato the possibility of a now or different kind of accident from any accident previously evaluated?

Operation of PNPS in accordance with the proposed changes will not croato the possibility of a now or different kind of accident from any accident previously evaluated because of the following:

The proposed more rostrictivo toquirements will not alter the plant configuration (no now or different type of equipment will be installed) or chango methods governing plant operation. The chango does imposo different requiroments. Howevor, the changes are consistont with assumptions mado in the safety analyses.

3. Does this change involvo a significant reduction in a margin of safety? ,

Operation of PNPS in accordance with the proposed changes will not involve a significant reduction in a margin of safety because of the following:

The proposed more rostrictive requirements will not alter assumptions relative to mitigation of an accident or transient event or alter the operation of process variables,

, structures, systems, or components as described in the safoty analyses.

2

__ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION

. CTS 3.5 l CORE AND CONTAINMENT COOLING SYSTEMS :

IECHNICAL CHANGES RELOCATIONS I' (Ri Labeled Discussion of Changes, CTS 3.5)

This change proposes to relocate the current pump and motor operated valve testing '

requirements (4.5.B.1.a and 4.5.B.1.b) to the PNPS IST program (Specification 3/4.13) and the procedures implomonting the IST program. Any changos to this testing provision will be adequately controlled by 10 CFR 50.55a and 10 CFR 50.59.

BECo has evaluated this proposed Technical Specification chango and has determined that it !

involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided.

i

1. Does the chango involvo a significant inaase in the probability or consequences of an accident previously evaluatod?

i Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously '

evaluated because of the following:

This proposed change relocates requirements from the Technical Specifications to the Inservice Testing (IST) Program. The (IST) Program documents containing the relocated requirements must bo maintained using the provisions of 10 CFR 50.55a and !

10 CFR 50.59. Sinco any changos to the (IST) Program documents will be evaluated i por 10 CFR 50.55a and 10 CFR 50.59, no increase in the probability or consequences of an accident previously evaluated will be allowed without NRC review. ;

2. Does the chango create the possibility of a new or different kind of accident from any accident previously ovaluated?

Operation of PNPS in accordance with the proposed chango will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: i This change relocatos requirements to the (IST) Program. This change will not alter the plant configuration (no new or different type of equipment will be installed) or changes in methods governing plant operation. This change will not imposo different requirements, and adequate control of Information will be maintained. This change will not alter assumptions made in the safoty analysis.

3. Does this change involvo a significant reduction in a margin of safety?

Operation of PNPS in accordance with the proposed change will not involve a signihcant reduction in a margin of safety because of the following: ,

This change relocates requirements from the Technical Specifications to the Inservico Testing (IST) Program. The requirements to be transposed to the IST program are the same as the existing Technical Specifications. Since any changos to the (IST)

Program documents will be evaluated per 10 CFR 50.55a and 10 CFR 50.59, no reduction in margin of safety previously approved will be allowed without NRC review.

3 J

. DETERi "iCN OF NO SIGNIFICANT HAZARDS CONSIDERATION CTS 3.5 CORE AND CONTAINMENT COOLING SYSTEMS i 311CWJCALCllAN0liSallSSJES1RIC11YE ,

(Li Labeled Discussion of Changes for CTS 3.5) l This chango proposes a now ACTION and Completion Time for one required RBCCW pump !

inoperable. With one required RBCCW pump inoperable, the remaining pump in the affected loop is sufficient to handle the normal operation heat loads, and the remaining OPERABLE j loop is sufficient to support the required actions of SPECIFICATIONS 3/4.5.E.1,

Suppression Pool Cooling System" and 3/4.5.B.2

Containment Spray System" for one subsystem I inoperablo. Tho 7 day completion time is consistent with the completion times for one Inoperable loop of suppression pool cooling system or containment spray system. The 7 days is adoquate for this reason and the low probability of an event occurring during this period that would requito RSCCW to support the core and containment cooling functions. ,

BECo has ovaluated this proposed Technical Specification change and has determined that it involvos no significant hazards consideration. This determination has been performed in accordance with the critoria set forth in 10 CFR 50.02. The following evaluation is provided.

1, Does the chango involve a significant incruaso in the probability or consequences of an i accident previously evaluated? ,

Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluatud because of the following:

This change relaxes the current requirements to declaro the affected RBCCW subsystem inoparable when one of the required RBCCW pumps is in:perable. Since the RBCCW system is not assumed as an Initiator of any analyzed event, the proposed change will not affect the probability of an accident occurring. The safety function of the RBCCW systom is to support the operability of the RHR suppression pool cooling i and spray functions, and component cooling for the RHR and core spray pumps, and area coolors. With one required RBCCW pump inoperable, the remaining pump in the affected subsystem is capable of supporting the component cooling requirements for !

the RHR and core spray pumps, and area coolers, and the remaining OPERABLE

< subsystem is capable of supporting the suppression pool cooling and spray functions.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following

The proposed change will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these systems are operated, maintained, modified, tested, or inspected.

4 1

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.- DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION ;

CTS 3.5 !

CORE AND CONTAINMENT COOLING SYSTEMS _ l i

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3. Does this change involve a significant reduction in a margin of safety? l I Opeiation of PNPS in accordance with the proposed change will no; hvolve a f

, signliicant reduction in a margin of safety because of the following:

,, [

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The 7 day completion time is consistent with the completion times f" one inoperable l

.1 loop of suppreselon pool cooling system or containment spray systum, and the l remaining pump in the affected subsystem is capable of supporting the component ;

cooling requirements for the RHR and core spray pumps, and area coolers. l 1 i i

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ATTACHMENT D :

I Revised Technical Specifications f

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LIMITING CONDITIONS FOR OPERATION $URVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4,5 QQRE AND CONTAINMENT COOLING SlSIFJdS SYSTEMS 6Rolicabilitv Aopticabilitv Applies to the operational status Applies to the Surveillance of the core and suppress!on pool Requirements of the core and cooling systems, suppression pool cooling systems which are required when the corresponding Limiting Condition for operation is in effect.

Qbje.,_qRye Obiective To assure the operability of the To venfy the operability of the core and suppression pool cooling core and suppression pool cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capability is an essential response to station an essential response to statioi, abnormalities. abnormalities.

Specifigation Specification A. Core Spigy and LPCI Systems A. Core Sorav and LPCI Systems

1. Both core spray systems shall 1. Core Spray System Testing.

be operable whenever irradiated fuelis in the vessel and prior item Freauency to reactor startup from a Cold Condition, except as specified in a. Simulated Once/

3.5.A.2 below, Automatic Operating Actuation Test. Cycle

2. From and after the date that one of the core spray systems is made or b. Pump When tested found to be inoperable for any Operability, as specified in reason, continued reactor 3.13 verify that operation is permissible during the 'ich core :

succeeding seven days, provided ray pump that during such seven days all oelivers at active components of the other least core spray system and active 3300 GPM l components of the LPCI system against a and the diesel generators are system head operable corresponding .

to a reactor l i

vessel pressure of l 104 psig.

1

)

1 Amendme..t No. 3/4.5-1 !

i LIMITING CONDITIONS FOR OPERATION SURVElLLANCE REQUIREMENTS i i

3.5 COBE AUD CONTAINMENT COOLING 4.5 CORE AND CONTAINMEET COOUNG '

SYSTEMS SYSTEMS 1 .

A. Core Spay _ add LPCI Systems (Cont) A. Core Spray and LPCl Systems (Cont) l

1. c. Motor As Specified Operated in 3.13 i Valve ,

Operability I

d. Core Spray Header Ap Instrumentation l

Check Once/ day l Calibrate Once/3 i months Test Step Once/3 !

4 months !

1

2. This section intentionally left blank

3. The LPCI system shall be operable 3. LPCI system testing shall be as whenever irrediated fuelis in the follows:

reactor vessel, and prior to reactor j startup from a Cold Condition, a. Simulated Once/

except as specified in 3.5.A,4 and Automatic Operating 3.5.F.5. Actuation Test. Cycle

4. From and after the date that the b. Pump When tested LPCI system is made or found to be Operability. as specif,ad 3

inoperable for any reason, continued in 3.13, verify reactor operation is permissible only that each during the succeeding seven days LPCIpump unless it is sooner made operable, delivers l

provided that during such seven 4800 GPM at days the active companents of both a head across core spray systems and the diesel the pump of generators required for operation of at least 380 such components,if no external _ __ ft.

source of power were available, shall be operable, c. Motor As Specified Operated in 3.13 1

5. ~ If the requirements of 3.5.A cannot Valve ;

- be met, an orderly shutdown of the Operability -

reactor shall be initiated and the l reactor shall be in the Cold l Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . i l

Amen'dment No. 3/4.52

- LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING S.YSTEMS SYSTEMS l B.1 Residual HeaLR_emoval(RHR) B.1 Residual Heat Refpoval(RHR)

Syppression Pool Coolina Suppression Pool Coolino Specif!ga_tiga Two RHR suppression pool cooling 1. Venfy each RHR suppression pool subsystems shall be OPERABLE. cooling subsystem manual, power operated, and automatic valve in 6pplicability: the flow pa;h that is not locked, sealed, or otherwise secured in Whenever irradiated fuelis in the reactor position is in the correct position or vessel, reactor coolant temperature is can be aligned to the correct

>212' F, and prior to startup from a cold position every 31 days.

condition.

2. Verify each RHR pump develops a g flow rate 2 5100 GPM through the associated heat exchanger while A. One RHR suppression pool cooling operating in the suppression pool subsystem inoperable, cooling mode as specified in Specification 3/4.13.

1. Restore the RHR suppression pool cooling subsystem to OPERABLE status within 7 days.

B. Required Action and associated Completion Time not met, 9.6 Two RHR suppression pool cooling subsystems inoperable,

1. Be in Cold Shutdown within

- 24 hou*s.

Amendment No. 3/4.5-3

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

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LIMITING CONDITIONS FOR OPERATION AWRyf&lJdGE. fig 2Wi!!CMENIE i 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONT *.INMENT COOLING !

SYSTEMS SYSTEMJ l B.2 Residual Heat Removal (RHR) 3.2 Residual Heat Removal (RHR) i

C201ainment Sorav containment Spray l I

i Specification:

. Two RHR containment spray 1. Venfy each RHR containment .

subsystems shall be OPERABLE. spray subsystem manual, power i operated, and automatic valve in

> 6pplicability: the flow path that is not locked, .

sealed, or otherwise secured in Whenever irradiated fuelis in the position it ni the correct position i reactor vessel, reactor coolant or can be aligned to the correct temperature is >212T, and prior to position every 31 days.

startup from a cold condition.

! 2. Air test drywell and suppression i Actions: pool (torus) headers and nozzles A. One RHR containment spray subsystem inoperable, i

1. Restore RHR containment i spray subsystem to OPERABLE status within 7 days.

1 B. Required Action and associated Completion Time not met 98 '

Two RHR containment spray subsystems inoperable, Be in Cold Shutdown within 1.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

L Amendment No. 3/4.5-4 y m e- m - yy,e>ue y yeghae .,-- .y-,--seayr>*,fyg.g .p e -tr,,..=p pg,ymv.,-9,.,,.-.,s,,q,.-m,gg-y,,,,w,,%g.pg, egy 'p+rp-.g.q.gc-39,,u-,gwy ,- ,,,.g%.u.e-,,3y-+eyy.g-+-.w-

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LIMITING CONDITIONS FOR OPERATION SURVElLLANCE REQUIREMENTS 3

3.6 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING l SYSTEMS SYSTEMS !

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B.3 Spa _qigr39il ding Closed Coolina Wa.ipf B.3 Reig. tor Buildina Closed Cqgh0 dater )

i (RBCCW) System (RBCCW)Jvstem i Specificatjo.D; i

! Two RBCCW subsystems shall be 1. ..~....-------- N O T E-------- ~ ~~-

l 1 OPERABLE. Isolation of flow to individual i 1

j components does not render the 6pR!G.apli l t tyi RBCCW subsystem inoperable.

. ~..... ..... . ... ...... . . . ..

Whenever irradiated fuelis in the reactor vessel, reactor coolant Verify each RBCCW manual, power ;

i operated, and automatic valve in the temperature is >212' F, and prior to ,

flow path, that is not locked, sealed, or startup from a cold condition,

otherwise secured in position, is in the ,

correct positien or can be aligned to tne i Actions:

correct position every 31 days. ;

i A. One required RBCCW Pump

inoperable,

1. Restore the required RBCCW pump to OPERABLE status within 7 days. ,

B. One RBCCW subsystem -

inoperable, 4

1. Restore the RBCCW '

subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

C, Required Action and associated Completion Time not met, i s

.j

Two RBCCW subsystems I

inoperable,

1. Be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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I Amendment No. 3/4.5-5 i

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' WMITING CONDIT60NS FOR OPERATION SURVEILLANCE REQUIREMENTS t 3.5 ff REA.NQ.QQNTAINMENT COOLING 4.0 CORE AND CONTAINMENT COOLING l SldTIMS SYSTEM,3 l B.4 Salt Service Water (SSW)1 Svatem B.4 Salt Service Water (SSW)1 System I

Specification

Verify the water levelin each

~

Two SSW subsystems shall be 1.

2

< OPERABLE. SSW pump well of the intake 4

structure is 213 ft 9 in below

! Applicapll.!1E mean sea level every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 Whenever irradiated fuelis in the 2. Verify the average sea water

) reactor vessel, reactor coolant temperature is s 75'F every 24 temperature is >212* F, and prior to hours.

i startup from a cold condition.

3. -- --~~ N OT E --- -- ---

Actions: Isolation of flow to individual :

i components does not render the !

A. One SSW subsystem inoperable, SSW subsystem inoperable. f

~ ~ " " " " " " " ~ ~ ~ ~ ~ " " " ~ ~ ~

1, Restore the SSW subsystem I to OPERABLE status within Verify each SSW subsystem 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . manual, power operated, and automatic valve in the flow paths i

. B. Required Action and associated servicing safety related systems or components, that is not locked, Completion Time not met, sealed, or otherwise secured in pg p sition, is in the correct position 4

Two SSW subsystems inoperabia, every 31 days.

4 9.8 UHS inoperable, 4. Verify each SSW subsystem actuates on an actual or l

1. Be in Cold Shutdown within simulateu initiation signal every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 2 years.

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t Amendment No. 3/4.5-6

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LIMITING CONDITIONS FOR OPERATION 111RVEILLANCE REQUIREMENTS 3.5 CORE AtQ CONTAINMENT COOLINS 4,5 . CORE AND CONTAINMENT COOLING l 3

SYSTE_M3 SYSTEMS j C. HPCI Sv1191D C. HPCI System ,

1. The HPCI system shall be 1. HPCI system testing shall be as operable whenever there is follows:

irradiated fuel in the reactor vessel, ,

i reactor pressure is greater than a. Simulated Once/ '

150 psig., and reactor coolant Automatic Operating i

temperature is greater than 365'F, Actuation Cycle

. except as specified in 3.5.C.2 Test 4 below.

b. Pump When tested

2. From and af ter th., date that the Operability as specified in

! HPCI system is made or found to 3.13, verify ,

. be inoperable for any reason, that the HPCI continued reactor operation is pump delivers ,

permissible only during the at least 4250 t succeeding 14 days unless such GPM for a system is sooner inade operable, system head providing that during such 14 days corresponding

all active components of the ADS to a system, the RCIC system, the reactor

LPCI system and both core spray pressure of systems are operable. 1000 psig.

l 3. If the requirements of 3.5.C cannot c. Motor As Specified in

be met, an orderly shutdown of the Operated 3,13 reactor shall be initiated and the Valve

. reactor shall be in the Cold Operability Shutdown Condition within -

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . d. Flow Rate at Once/

, 150 psig. operating cycle, venfy that the HPCI pump delivers at least 4250 GPM foi a system head corresponding to a reactor pressure of 150 psig.

The HPCI pump shall deliver at least 4250 GPM for a system head corresponding to a reactor pressure of

- 1000 to 150 psig.

-i

( l

)

Amendment No. 3/4.5 7 l

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

WMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOUNS 4.5 CORE AND CONTAINMENT COOLING l ESIEMS SYSTEMS ,

D. Reactor Core Isolation Coolina D. Reactor Core isolation Coolina (RCIC)

(RCIC) System System

1. The RCIC system shall be 1. HPCI system testing shall be as operable whenever there is follows:

Irradiated fuel in the reactor vessel, '

reactor pressure is greater than a. Simulated Oncel

150 psig, and reactor coolant Automatic Operating temperature is greater than 365'F, Actuation Test Cycle except as specified in 3.5.D.2 below. b. Pump When tested as Operability specified in

2. From and after the date that the 3.13, verify that RCIC systern is made or found to the RCIC pump be inoperable for any reason, delivers at least continuod reactor operation is 400 GPM at a !

permissible only during the system head !

succeeding 14 days unless such corresponding system is sooner made operable, to a reactor providing that during such 14 days pressure of the HPCIS is operable. 1000 psig.

3. If the requirements of 3.5.D cannot c. Motor As Specified be met, an orderly shutdown of the Operated in 3.13 reactor shall be initiated and the Valve reactor shall be in the Cold Operability Shutdown Condition within l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . d. Flow Rate at Once/

150 psig. operating cycle verify that the RCIC pump delivers at least 400 GPM at a l system head -

corresponding to a reactor pressure of 150 psig.

The RCIC pump shall deliver at least 400 GPM for a system head corresponding to a reactor pressure of 1000 to 150 psig.

Amendment No. 3/4.5-8 ,

1 LIMITING CONDITIONS FOR OPERATION SURVElLLANCE REQUIREMENTS 3.5 CQRE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS r i E. 6.u. tomatic Depressurization Svilem E Automatic Qqoressurization Svilem l (ADS) (ADS)

1. The Automatic Depressurization System shall be operable ,

whenever there is irradiated fuelin 1. During each operating cycle the

the reactor vessel and the reactor following tests shall be performed -

pressure is greater than 104 psig on the ADS: r i and prior to a startup from a Cold l Condition, except as specified in a. A simulated automatic 3.5.E.2 below, actuation test shall be

] performed prior to startup after

2. From and after the date that one each refueling outage.

valve in the Automatic The ADS manualinhibit switch Depressurization System is made will be included in this test.

or found to be inoperable for any !

reason, continued reactor b. With the reactor at pressure, 1

operation is permissible only each relief valve shall be

=

during the succeeding 14 days manually opened until a ,

i unless such valve is sooner made corresponding change in operable, provided that during reactor pressure or main such 14 days the HPCI system is turbine bypass valve positions operable- indicate that steam is flowing

3. If the requirements of 3.5,E cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i r

4 2

N b Amendtr.ent No. 3/4.5-9 l

4 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 35 QQ.Rii AND CONTAINMENT CQQLING 4.5 CORE AND CONTAINMENT QOOLING -

S.YSlEMS S.Y.SIEMS l F. Minimum Low Pressure Coolina and F. Minimum Low Prepure Coolina and QLesel Generator AvailabMy Diesel Generator Availability 1 1 During any period when one diesel 1. When it is determined that one generator is inoperable, continued diesel generator is inoperable, reactor operation is permissible only within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , determine that the dunng the succeeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months operable dietel generator is not unless such diesel generator is inoperable due to a common sooner made operable, provided cause failure, that all of the low pressure core and containment cooling systems and 9I the remaining diesel generator shall be operable. If this requirement II (9Ma for cannot be met, an orderly shutdown the operable diesel generator, shall be initiated and the reactor shall be placed in the Cold and Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2 Any combination of inoperable 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaf ter, verify correct components in the core and breaker alignment and indicated containment cooling systems shall power for each offsite circuit.

not defeat the capability of the remaining uperable components to fulfill the cooling functions.

3 When irradiated fuel is in the reactor vessel and the reactor is in the Cold Shutdown condition, both core spray systems, and the LPCI system may be inoperable, provided no work is being done which has the potential for draining the reactor vessel.

4 During a refueling outage, for a period of 30 days, refueling operation may continue provided that one core spray system or the LPCI system is operaole or Specification 3.5.F.5 is met.

Amendment No. 3/4.5 10 l

. t a

LIMITING CONDITIONS FOR OPERATION SURVElLLANCE REQUIREMENTS 3.5 CORE ANQCONTAINMENT COOLING 4.5 QQEE AND CONTAINMENT COOLING j SYSTEMS SYSTEMS ;

F Minimum Lgw Pressure Coolina an.d F. Minimum Low Pressure Coolina and

, pitselGenttajgLA_yJjlatbAtyA (Cont) Q3felGenerator Availabilitv (Cont) i i *

5. When irradiated fuelis in the reactor vessel and the reactoris in i

the Refueling Condition with the torus drained, a single control rod ,

drive mechanism may be removed, l if both of the following conditions i

are satisfied: ,

a) No work on the reactor vessel, in addition to CRD removal, wdl be performed which has the potential for ,

exceeding the maximum :

leak rate from a single control blade seat if it became unseated.

b) i) the core spray systems are i operable and aligned with a suction path from the condensate storage tanks.

li) the condensate stoage ,

tanks shall contain at least i 200,000 gallons of usable water and the refueling cavity and dryer / separator pool shall be flooded to a least elevation 114' 0" i

G. Deleted i

Amendment No, 3/4.5 11

. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 35 QQHEf3D_ CONTAINMENT COOLING 45 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS Maintename_of Filled Dischar.2e Pipe H. Maintenance of Filled Discharoe Pipe H. {

Whenever core spray systems, LPCI The following surveillance requirements assure that the system, HPCI or RCIC are required to be operable, the discharge piping from discharge piping of the core spray l

. the pump discharge of these systems to sys'. ems, LPCI system, HPCI and r the last block valve shall be filled. RCIC are filled:

.i !

t

1. Every month the LPCI system l and core spray system discharge piping shall be vented from the t high point and water flow observed.

2. Following any period where the .

LPCI system or core spray ;

i systems have not been required

to be operable, the discharge ,

piping of the inoperable system ,

i shall be vented from the high point prior to the return of the system to service, i 3. Whenever the HPCI or RCIC l system is lined up to take suction l from the torus, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

f a

4 i

4 4

Amendment No. 3/4.5 12 1

-- .~#,,, -.,< - , - - ,-. . - _ . , -,- - . ., --,m,.. ,--, .. _ . - - _ . - . ,--...,,--.~m,,.-.. . , .--. . - - ~ ~ . - ,, - - - - - - -.v-,_..

\

1

}

ATTACHMENT E Revised Bases

- Cere Spray and LPCI Syst:m B 3/4.5.A B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS

- B 3/4.5.A Coro Spray and L"Cl System i BASES:

, BACKGROUND Each Core Spray system consists of one pump and associated piping and valves with all active components required to be operable. The

- LPCI system consists of four LPCl pumps and associated piping and valves with all active components required to be operable.

l The LPCI system is not considered inoperable when the RHR System is operating in the shutdown cooling mode.

5 APPLICABLE Based on the loss of coolant analysis performe by General Electric in SAFETY ANALYSES accordance with Section 50.46 and Appendix K of 10CFR50, the-Pilgrim i Emergency Core Cooling Systems are adequate to provide sufficient cooling to the core to dissipate the energy associated with

the loss of coolant accident, to limit calculated fuel clad temperature to less than 2200' F, to limit calculated loca! metal water reaction to less >

than or equal to 17%, and to limit calculated core wide metal water reaction to less than or equal to 1% The detailed bases is described in NEDC 31862P and summarized in Section 6.5 of the PNPS FSAR.

d The analyses discussed in NEDC 31852P calculated a peak clad fuel temparature of less than 2200'F with a Core Spray pump flow of 3200 gallons per minute (gpm). A flow rate of 3300 gpm eiisures adequate ;

flow fur events involving degraded voltage.

i Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Pilgrim, to exceed the minimum requirements by at least 25%. In addition, cooling effectiveness has been demonstrated at less than half the rated ficw in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis takes credit for core spray flos inh the core at vessel pressure below 205 psig. However, the anaijsis is conservative in that no credit is taken for spray cooling heat transfer in the hottest fuel bundle until the pressure at rated flow for the core spray (104 psig vessel pressure)is reached.

The LPCI system is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system actions in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI system and the core spray system provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance from the high pressure emergency core cooling systems. The analyses in NEDC-31852P calculated a peak clad fuel temperature of less than 2200 F with LPCI pump flows of 4550 gpm,4033 gpm, and 3450 gpm for two, three, and four pump combinations feeding into a single loop. A single pump flow rate at 4800 gpm ensures sufficient flow to meet or exc9ed the analyses' assumptions.

(continued)

Amendment No. B3/4.5-1 l

. 4 o Ccre Spray and LPCI Syst:m B 3/4,5.A

~

BASES j APPLICABLE The analyses of LOCA for PNPS demonstrated the combination of I SAFETY ANALYSES LPCS/LPCI systems are sufficient to provide core cooling even with a (continued) ' singie failure of either an active or passive safety related component.

The analyses determined there were four significant single failures that challenge the Emergency Core Coolant Systems' capability to prevent fuel damage during the postulated LOCA. They are:

1) Battery Failure Loss of a single battery train could leave only one LPCS pur10, two LPCI pumps, and ADS to mitigate the LOCA ?his is '.ne mo:t limiting single failure for all but the largest piistulated recirculation line breaks and for all postulated ron-recirculation line breaks.

2) LPCI Injection Valve Failure - Loss of the injection valve

' selected by LPCI Loop Selection Logic for the pathway for all LPCI pumps' flow leaves two core spray pun'ps, HPCI, and ADS for LOCA mitigation. This becomes the limiting single failure for the largest postulated recirculation line breaks.

3) Loss of one emergency diesel generator This leaves one LPCS pump, two LPCI pumps, and ADS for LOCA mitigation.

4) HPCI Failure This leaves all other ECCS resources 4

available. It is a significant failure primarily for smallline breaks.

I in all cases above, the remaining ECCS resources are sufficient to prevent PCT from eneeding 2200 F and other criteria provided in Section 50.46 and Appendix K of 10CFR50.

ACTIONS Should one Core Spray system become inoperable, the remaining Core Spray ar,d the LPCI system are available should the need for core cooling arise. Based on judgments of the reliability of the remaining systems (i.e., the Core Spray and LPCI) a seven-day repair period was obtained, SURVEILLANCE The testing interva! for the core and containment cooling systems is REQUIREMENTS based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation.. To increase the availability of the core and containment cooling systems,. the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated va%s are tested in

. accordance with ASME B&PV Code,Section XI (lWP and IWV, except where specific relief is granted) to assure their operability. The frequency and methods of testing are described in the PNPS IST l

Amendment No. B3/4.5 2 !

t 4 . ,

Cere Spray and LPCI System B 3/4,5.A -!

BASES

~

SURVEILLANCE - program. The PNPS IST Program is used to assess the operational-REQUIREMENTS readiness of pumps and valves that are safety related or important to (continued). safety. When components are tested and found inoperable the impact on system operability is determined, and corrective action or Limiting Conditions of Operation are initiated. A simulated automatic actuation i- test once each cycle combined with code inservice testing of the pumps and valves is deemed to be adequate testing of these systems.

The surveillance requirements provide adequate assurance that the core and containment cooling syst9ms will be operable when requded.

k Amendment No. B3/4.5-3 b

' RHR Suppressi::n Pocl Cocling B 3/4.5.B.1 B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS B 3/4.G.B.1 Residual Heat Removal (RHR) Suppression Pool Cooling BASES __

BACKGROUND Following a Design Basis Accident (DBA), the RHR suppression pool cooling subsystem removes heat from the suppression pool. The suppression poolis designed to absorb the sudden input of heat from the primary system. In the long term, the pool continues to absorb residual heat generated by fuelin the reactor core. Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by two redundant RHR suppression pool cooling subsystems. The purpose of this Specification is to ensure that both subsystems are OPERABLE in applicable MODES.

Each RHR suppression pool cooling subsystem contains two pumps and one heat exchanger and is manually initiated and independently controlled. The two subsystems perform the suppression pool cooling function by circulating water from the suppression pool through the RHR heat exchangers and returning it to the suppression pool. The RHR heat exchangers (tube side) are cooled by the reactor building closed cooling water system (Specification 3/4.5.B.3), which is in turn cooled by the salt water service system (Specification 3/4.5.B.4)

S/RV leakage and high pressure coolant injection or reactor core isolation cooling system testing increase suppression pool temperature more slowly. The RHR suppression pool cooling subsystem is also used to lower the suppression pool water bulk temperature following such events.

APPLICABLE Reference 1 contains the results of analyses used to predict primary SAFETY ANALYSES containment pressure and temperahre following large and small break LOCAs. The intent of the arialyses is to demonstrate that the heat removal capacity of the RHR suppression pool cooling subsystem is adequate to maintain the primary containment conditions within design limits. The suppression pool temperature is calculated to remain below the design limit.

The RHR suppression pool cooling subsystem satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

(continued)

Amendment No. B1M.5-4

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

4' RHR Suppressi:n Pecl Cotling B 3/4.5.B,1 B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES 4

SPECIFICATION During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref.1). To ensure that these requirements are met, two redundant suppression pool cooling subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR suppression pool cooling subsystem is OPERABLE when two RHR pumps, the heat

( exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.

APPLICABILITY When reactor coolant temperature is > 212*F and irradiated fuelis in the reactor vessel, a DBA could cause a release of radioactive !

material to primary containment and cause a heat up and :

pressurization of pnmary containment. When the reactor temperature is s 212 F , the probability and consequences of these events are less severe. Therefore, the RHR suppression pool cooling subsystem is not required to be OPERABLE when reactor coolant temperature is s 212 F.

. ACTIONS J A

With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a

. single active failure in the OPERABLE subsystem could result in reduced primary containment cooling capability The 7 day completion time is acceptable in light of the redundant RHR suppression pool cooling capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.

k.1 If the inoperable RHR suppression pool cooling subsystem cannot be restored to OPERABLE status within the associated completion time or if two RHR suppresr. ion pool ecoling subsystems are inoperable, the plant must be brought to a condition in which the specification

.does not apply. To achieve this status, the plant must be brought to Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The all owed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued) .

- Amendment No. B3/4.5-5

o RHR Suppressitn Potl Cscling B 3/4.5.B.1 B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES SURVEILLANCE SR 4.5 B.1.1-REQUIREMENTS 1

Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow l path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the noneccident position provided it can ,

be aligned to the accident position wi%n the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any ,

testing or valve manipulation; rather, it involves verification that those

. valves capable of being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

t 4

The frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a

! single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This

frequency has been shown to be acceptable based on operating expenence.

s i SR 4,5.B.1.2 4 Verifying that each RHR pump develops a flow rate 2 5100 gpm (Ref.1) while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that pump performance has not degraded during the cycle. Flow is a normal

' test of centrifugal pump performance required by ASME Code,Section XI (Ref. 2). This test confirms one point on the pump design curve, and 'he results are indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The frequency of this SR is in accordance with the Inservice Testing Program, Specification 3/4.13.

REFERENCES 1. FSAR, Section 14.5.

2.- ASME, Boiler and Pressure Vessel Code,Section XI.

Amendment No. 83/4.5-6 i

RHR Centainment Spray B 3/4.5.B.2 B 3/4.5 - CORE AND CONTAINMENT COOLING SYSTEMS B 3/4.5.B.2 Residual Heat Removal (RHR) Containment Spray BASES BACKGROUND The RHR containment spray subsystem provides containment spray capability as an alternate method for reducing containment pressure (temperature) fo!!owing a LOCA. A portion of the water pumped through the RHR heat exchangers can be diverted to spray headers in the drywell and above the suppression pool. The portion of the RHR heat exchanger flow to the spray headers in the drywell condenses any steam that may exist in the drywell thereby lowering containment pressure (temperature). The remaining portion returns to the suppression pool via the suppression pool bypass liner The spray collects in the bottom of the drywell until the water level rises to the level of the pressure suppression vent pipes where it overflows and drains back to the suppression pool. Approximately 5 percent of the total flow may be directed to the suppression chamber spray ring '

4 to cool any noncondensable gases collected in the free volume above the suppression pool. The containment spray subsystem will remove

- energy from the drywell by condensing steam, thereby, making available the drywell volume to accommodate additional quantities of gases from any postulated metal water reactions above that which

- the containment can inherently accommodate without spray. (Ref.1)

The containment spray mode of the RHR cannot be operated unless the levelinside the reactor vessel shroud is above the two thirds core

height set point and the drywoll pressure exceeds a setpoint greater than 1 but less than 2 psig.

Interlocks are provided to prevent LPCI flow from being diverted to the containment spray mode unless the core is flooded. A keylock switch in the control room permits the overriding of this interlock to reduce containment pressure if required.

Each of the two RHR containment spray subsystems cor;tains two pumps and one heat exchanger, which are manually initiated and indepandently controlled. The RHR heat exchangers (tube side) are cooled by the reactor building closed cooling water sy , tem (Specification 3/4.5.B.3), which is in tum cooled by the salt service water system (Specificabn 3/4.5.B.4). Either RHR containment spray subsystem is sufficient to condense the steam from small bypass leaks hom the drywell to the suppression chamber airspace during the postulated DBA.

(continued)

Amendment No. B3/4.5-7

i . RHR Centainment Spray B 3/4,5.B.2

^

B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES APPLICABLE Reference 2 contains the results of analyses used to predict primary -

SAFETY ANALYSES containment pressure and temperature following loss of coolant t accidents. The intent of the analyses is to demonstrate the pressure

! (temperature) reduction capability of the RHR containment spray :

system.

The RHR containment spray subsystem satisfies Criterion 3 of 10 CFR 50.36(c)(2)(li).

i SPECIFICATION in the event of a DBA, a minimum of one RHR containment spray subsystem is required to mitigate potential bypass leakage paths and maintain the primary containment peak temperature below the design limits (Ref 3). To ensure that these requirements are met, two redundant RHR containment spray subsystems must be OPERABLE 3 with power from two safety related independent power supplies. l 4

Therefore, in the event of an accident, at least one subsystem is

OPERABLE assuming the worst case single active failure. An RHR containment spray subsystem is OPERABLE when one of *he pumps, the heat exchanger, and associated piping, valves, instrumentation, e

and controls are OPERABLE.

l APPLICABILITY When reactor coolant temperature is > 212 F and irradiated fuelis in the reactor vessel, a DBA could cause a release of radioactive material to primary containment and cause a heat up and pressurization of primary containment. When the reactor temperature is s 212 'F, the probability and consequences of these everns are less severe. Therefore, the RHR containment spray system is not

required to be OPERABLE when reactor coolant temperature is s 212*F.

ACTIONS A_J With one RHR containment spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within

+

7 days. In this condition, the remaining OPERABLE RHR containment spray subsystem is adequate to perform the primary containment bypass leakage mitigation function. However, the overall reliability is reduced because a single active failure in the OPERABLE subsystem could result in reduced primary containment peak temperature control. The 7 day completion time was chosen in

- light of the redundant RHR containment spray capabilities afforded by the OPERABLE subsystem and the low probability of a LOCA occurring during this period.

(continued)

Amendment No. B3/4.5-8

L

- RHR Centainm:nt Spray B 3/4,5.B,2 B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES ACTIONS EL1

- (continued)

If the inoperable RHR containment spray subsystem cannot be restored to OPERABLE status within the associated completion time or if two RHR containment spray subsystems are inoperable, the plant must be brought to a condition in which the .oecification does not apply. To achieve this status, the plant must be brought to Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

S'JRVEILLANCE SR 4.5.B 2.1 REQUIREMENTS Verifying the correct alignment for manual, power-operated, and automatic valves in the RHR containment spray mode flow path provides assurance that the proper flow paths will exist for system operation. This SR does riot apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This frequency has been shown to be acceptable based on operating experience.

SR 4.5.8.2.2 Verifying that the drywell and suppression pool (torus) headers and nozzles are free 'of obstructions by blowing air through them ensures an open flow path. The frequency for performance of the spray -

nozzle obstruction surveillance test of 5 years is justified due to the passive design of the nozzles and has been shown acceptable through industry operating experience.

(continued)

Amendment No. 83/4.5 9

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

- RHR C:ntainment Spray -

B 3/4.5 B.2 B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS -

BASES

-REFERENCES 1. FSAR, Section 4.8

2. FSAR, Section 14.5.

3. ASME, Boiler and Pressure Vessel Code,Section XI.

I

' Amendment No. B3/4.5-10 i

4

ROCCW Syst:m B 3/4.5.B.3 B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.B.3 Reactor Building Closed Cooling Water (RBCCW) System

, BASES

~

BACKGROUND The RBCCW system is designed to provide a heat sink for the Residual Heat Removal (RHR) system heat exchangers and the removal of heat from emergency core cooling system (ECCS) equipment, such as RHR pump lube oil coolers, core spray pump

. motor thrust bearings, and room coolers, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient.

{

The RBCCW system consists of two full capacity closed loops. Each loop has three centrifugal pumps, rated 1,700 gal / min at 100 ft total dynamic head (TDH), taking suction fror, the reactor building cooling -

water heat exchanger and capable of delivering inhibited demineralized water to the associated equipment. In order to transfer the design RHR system heat load during postulated transient or 1 accident conditions, a minimum of two pumps in one loop and a cooling water heat exchanger is required.

Motor operated gate valves are provided in each loop to manually isolate non-essential cooling loads under accident conditions. The two independent loops have the capability to be interconnected through two 12 inch cross ties. The valycs in the crossties are normally closed.

The RBCCW system is designed with sufficient redundancy so that no single active component failure can prevent it from achieving its design function. The RPCCW system is described in the FSAR, Section

10.5.5 (Ref.1).

!C During normal power operation, one pump in each loop is operating providing coolant flow to all of the associated equipment except the RHR heat exchangers which are valved off. Following a postulated loss of coolant accident (LOCA) coincident with loss of the preferred (offsite) AC power source, the operating RBCCW pumps will trip. One RBCCW pump in each loop is automatically restarted on its respective

,--- diesel generator approximately 30 sce after AC power is restored to the emergency service bus.

APPLICABLE The RBCCW system provides adequate cooling of safety equipment SAFETY ANALYSES required for safe reactor shutdown and removes heat from the suppression pool to limit the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary

containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RBCCW system to support long term cooling of the reactor or primary containment is discussed in the FSAR, Section 10.5.5.3 i- and 14.5.3.1.2 (Refs. 2 and 3, respectively). These analyses explicitly assume that the RBCCW system will provide adequate cooling support to the equipment required for safe shutdown. These analyses include

- the evaluation of the long term primary containment response after a design basis LOCA. (continued)

Amendment No. B3/4.5-11

C RBCCW Syst:m B 3/4.5.B.3 8 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES APPLICABLE The RBCCW System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

SAFETY ANALYSES (continued)

SPECIFICATION Two RBCCW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure j occurs coincident with the loss of offsite power. I An RBCCW subsystem is considered OPERABLE when:

1

a. Two pumps are OPERABLE; and i

b. An OPERABLE flow path is capable of taking suction from the t reactor building cooling water heat exchanger and transferring $

the water to the associated safety equipment and RHR heat exchangers at the assumed flow rate. Additionally, the RBCCW cross tie valves (which allow the two RBCCW loops to be .

connected) must be closed so that failure of one subsystem will l not affect the OPERABILITY of the other subsystems. l The isolation of the RBCCW system to individual components may render those components inoperable but does not affect the OPERABILITY of the RBCCW system.

! APPLICABILITY In all Modes except Cold Shutdown, the RBCCW system is required to .

be OPERABLE to support the OPERABILITY of the components or !

systems serviced by the RBCCW system, in Cold Shutdown, the OPERABILITY requirements of the RBCCW system are determined by the systems it supports.

f ACTIONS A_1 4

With one required RBCCW pump inoperable, the inoperable pump must be restored to OPERABLE status within 7 days. The remaining, required pump in the affected loop is sufficient to handle the normal operation heat loads and the remaining OPERABLE loop (2 required pumps) is sufficient to perform the RBCCW heat removal function.

However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced RBCCW capability.

The 7 day completion time is based on the remaining RBCCW heat )

removal capability and the low probability of a DBA with concurrent worst case single failure. !

l (continued) l

)

1 Amendment No. B3/4.5-12 l

- R::CCW Syst2m ,

B 3/4.5.B.3 l B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES ACTIONS ga ,

(continued) i With one RBCCW subsystem inoperable, the inoperable subsystem must be restoind to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . With the unit l in this condition, the remaining OPERABLE RBCCW subsystem is j adequate to perform the RBCCW containment cooling heat removal 1 function. However, the overall reliability is reduced because a single j active failure in the OPERABLE subsystem could result in loss of the RBCCW funcuon. The completion time is based on the capabilities afforded by the redundant OPERABLE RBCCW subsystem, the long term dependency of the RHR and Core Spray pumps for core cooling capability, and the low probability of an event occurring during this period requiring RBCCW.

i Q.1 I, t ..toperable RBCCW subsystem cannot be restored to OPERABLE status within the associated completion time or two RBCCW subsystems are inoperable, the unit must be placed in a MODE in which the Specification does not apply. To achieve this status, the unit must be placed in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The allowed completion times are reasonable, based on operating experience, to reach the required unit conditions from full power -

conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 4.5 B 3.1 REQUIREMENTS The 31 day frequency is based on engineering judgment, is consistent with the procedural controls goveming valve operation and ensures correct valve positions.

REFERENCES 1. FSAR, Section 10.5.5.1. ;

2. FSAR, Chapter 10.5.5.3.

3. FSAR, Chapter 14.5.3.1.2.

Amendment No. B3/4.5-13

.____j

SSW Syst:m B 3/4,5.8.4 B 3/4.5 CORE AND CONTAINMENT COCLING SYSTEMS 3/4.5.B.4 Salt Service Water System (SSW) and Ultimate Heat Sink (UHS)

BASES BACKGROUND The SSW system is designed to function as the ultimate heat sink for all the systems cooled by the Reactor Building Closed Cooling Water (RBCCW) and Turbine Building Closed Cooling Water (TBCCW) system during all planned operations in all operating states by continuously providing adequate cooling water flow to the se ondary sides of the RBCCW and TBCCW heat exchangers.

The SSW system consists of two open loops. Each loop has two pumps (plus a common spare) rated at 2,700 gal / min at 55 ft TDH, piping, valving, instrumentation, and controls as necessary to provide coolant to one RBCCW heat exchanger and one TBCCW heat exchanger on each loop. The pumps take a suction from Cape Cod -

Bay (UHS) and discharge to a common header from which independent piping supplies each of the two cooling water loops, each loop consisting of one reactor building and one turbine building cooling water heat exchangers. The water then returns to the bay from the outlet of the heat exchanger. Two division valves are included in the common discharge header to permit the SSW System to be operated as two independent loops. Either of the two subsystems is capable of providing the required cooling capacity (4500 gpm) to support the required systems with two pumps operating. The SSW system is described in FSAR, Section 10.7.5 (Ref.1).

During normal power operations,3 pumps are all that is necessary to supply cooling water to accommodate the heat load. To ensure that sufficient seawater flow is maintained through the RBCCW heat exchangers (minimum of 4500 gpm for each heat exchanger), motor-operated butterfly valves on the TBCCW heat exchanger outlets will automatically adjust to preset throttling positions and the RBCCW outlet valves will simultaneously open. Automatic adjustment of the outlet valves occur following a loss of coolant accident (LOCA) with a coincident Loss of Offsite Power (LOOP), or a LOCA with degraded voltage on the safety buses while being supplied from the startup transformer. If a LOCA occurs without a LOOP or ciegraded voltage condition, the heat exchanger outlet valves remain as-is. Manual adjustments of the outlet valves will be made by operators to achieve adequate cooling water flow.

A loss of AC power will trip all service water pumps and will close one of the two division valves in the common pump discharge header, effectively dividing the service water system into two independent loops. A selector switch determines which division valve will close and to which train the 'C" SSW pump will be dedicated on loss of AC power. Two pumps would be connected to each loop. The two division valves are arranged to permit the fifth (middle) pump to be operated on either loop.

(continued)

Amendment No. 83/4.5 14

SSW Syst;m B 3/4.5.B.4 B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES APPLICABLE The SSW system provides a supply of cooling water to the secondary ;

SAFETY ANALYSES side of the RBCCW heat exchangers adequate for the requirements of ;

the RBCCW under transient and accident conditions. The ability of the SSW system to support long term cooling of the reactor containment is assumed in evaluations of the equipment required for safe reactor shutdown presented in the FSAR, Section 10.5.5.3 and 14.5.3.1.2 (Refs. 2 and 3, respectively). These analyses include the '

evaluation of the long term primary containment response after a design basis LOCA.

The long term cooling capability of the RHR, core spray, and reactor building closed cooling water pumps is dependent on the cooling provided by the SSW system.

The SSW system, together with the UHS, satisfy Criterion 3 of e 10 CFR 50.36(c)(2)(ii).

SPECIFICATION The SSW subsystems are independent of each other to the degree that each has separate controls, power supplies, and the operation of one does not depend on the other, in the event of a DBA, one subsystem of SSW is required to provide the minimum heat removal capabllity assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, two subsystems of SSW must be OPERABLE. At least one subsystem will operate, if the worst single active failure occurs coincident with the loss of offsite power.

A subsystem is considered OPERABLE when it has an OPERABLE UHS, two OPERABLE pumps with associated controls and instrumentation and the following valves on that subsystem operabla:

1. One TBCCW heat exchanger outlet valve unless the valve is throttled.

2. Ono RBCCW heat exchanger outlet valve unless the valve is open.

3. One discharge header valve (i.e., the one opposite to the selected train) unless the valve is fully closed.

The OPERABILITY cf the UHS is based on having a minimum water levelin the pump well of the intake structure of > 13 ft 9 inches below mean sea level and a maximum water temperature of 75 F.

(continued)

Amendment No. B3/4.5-15

1 I

l

SSW Syst:m B 3/4.5.B.4 8 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES APPLICABILITY In all Modes except Cold Shutdown, the SSW system and UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the SSW systom.

In the Cold Shutdown Mode, the OPERABILITt' requirements of the SSW system and UHS are determined by the systems they support.

ACTIONS _A_1 With one SSW subsystern inoperable, the inoperable subsystem must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . With the unit in this condition, the remaining OPERABLE SSW subsystem is adequate to perform the RBCCW heat removal function. However, the overall reliability is reduced because a single active failure in the OPERABLE subsystem could result in loss of the SSW function. The completion time is based on the capabilities afforded by the redundant, OPERABLE SSW subsystem, the low probability of an event occurring during this period requiring SSW, and is consistent with the allowed Completion Time for restoring an inoperable RBCCW subsystem ,

!L1 If the inoperable SSW subsystem cannot be restored to OPERABLE status within the associated completion time or if two SSW subsystems are inoperable, or the UHS is inoperable, the unit must be placed in a MODE in which the Specification does not apply. To achieve this status, the unit must be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The allowed completion times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 4.5.B 4.1 REQUIREMENTS This SR verifies the water level in each pump well of the intake structure to be sufficient for the proper operation of the SSW pumps (net positive suction head and pump vortexing are considered in determining this, limit). The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency is based on operating expcrience related to trending of the parameter and the availabiSty of alarms to alert the operators prior to exceeding the limit.

SR 4.5.B.4.2 Verification of the sea water inlet temperaturo ensures that the heat removal capability of the SSW system is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to trending of the parameter and the availability of alarms to alert the operators prior to exceeding the limit.

(continued)

Amendment No. B3/4.5-16

}

SSW Syst3m i B 3/4.5.B.4 ,

B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES

- SURVEILLANCE SR 4.5.8.4.3 REQUIREMENTS (continued) Verifying the correct alignment for each manual, power-operated, and automatic valve in each SSW subsystem flow path provides assurance that the proper flow paths will exist for SSW operation.- This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A_ valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be automatically realigned to its accident position within the required time.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned [t are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation and ensures correct valve positions.

SR 4.5.B 4.4 This SR verifies the automatic adjustment of the motor operated butterfly valves on the TBCCW heat exchanger outlets, simultaneous opening of the RBCCW outlet valves, and closure of one of the two division valves in the connon SSW pump discharge header during an accident event. This is meanstrated by the use of an actual or simulated initiation signal. % SR also verifies the automatic start capability of one of the two SSW pumps in each subsystem.

The 24 month frequency of the Surveillance is based on engineering judgment taking into consideration the plant conditions required to perform the Surveillance, and is consistent with fuel cycle lengths.

REFERENCES 1. FSAR, Chapter 10.7.5

2. FSAR, Chapter 10.5.5.3

3. FSAR Section 14.5.3.1.2 Amendment No. B3/4.5-17

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

. HPCI Syst:m B 3/4.5.C B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.C. High Pressure Coolant injection (HPCI) System BASES Background HPCI is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss-of coolant which does not result in rapid depressurization of the reactor vessel. HPCI permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventcry until the vesselis depressurized. HPCI continues to operate until reactor vessel pressure is below the pressure at which LPCI l operation or Core Spray System operation maintains core cooling.

The capacity of the system is selected to provide this required core cooling. The HPCI pump is designed to pump 4250 gpm at reactor pressures between 1100 and 150 psig. Two sources of water are 4 available. Initially, demineralized water from the condensate storage ;

tank is used instead of injecting water from the suppression pool into the reactor.

When the HPCI System begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI System. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reached equilibrium with the flow through the break. Continued depressurization causes the break flow to decrease below the HPCI flow and the liquid inventory begins to rise. This type of response is typical of the small breaks. The core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity range of the HPCI.

APPLICABLE The limiting conditions for operating the HPCI System are derived from SAFETY ANALYSIS the Station Nuclear Safety Operational Analysis (FSAR Appendix G) and a detailed functional analysis of the HPCI System (FSAR Section 6).

l_ SPECIFICATION The requirement that HPCI be operable when reactor coolant temperature is greater that 365 F is included in Specification 3.5.C.1 to clanfy that HPCI need not be operable during certain testing (e.g.,

reactor vessel hydro testing at high reactor pressure and low reactor coolant temperature). 365 F is approximately cqual to the saturation steam temperature at 150 psig.

ACTION The analysis in FSAR Appendix G shows that the ADS provides a single failure proof path for depressurization for postulated transients and accidents. The RCIC is required as an alternate source of makeup to the HPCI only in the case of loss of all offsite AC power.

Considering the HPCI and the ADS plus RCIC as redundant paths,

_ and considering judgments of the reliability of the ADS and RCIC systems, a 14 day allowable repair time is specified. (continued) 4 Amendment No. B3/4.5-18

i.

l >

HPCI Syst:m B 3/4.5.C i

B 3/4,5 CORE AND CONTAINMENT COOLING SYSTEMS BASES ~ ,

' SURVEILLANCES The testing interval for the core and containment cooling systems is

' ased on industry practice, quantitative reliability analysis, judgment u !

and practicality. The core cooling systems have not been designed to be fully testable during operation. For example,in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable. To increase the availability of the core and containment cooling systems, the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested frequently, The pumps and motor operated valves are tested in accordance with ASME B&PV Code,Section XI (IWP and IWV, except where specific relief is granted) to assure their operability. The frequency and methods of testing are described in the PNPS IST program. The PNPS IST Program is used to assess the operational readiness of pumps and valves that are ;

i safety-related or important to safety. When components are tested and found inoperable the impact on system operability is determined, and corrective action or Limiting Conditions of Operation are initiated.

A simulated automatic actuation test once each cycle combined with code inservice testing of the pumps and valves is deemed to be adequate testing of these systems.

The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when :equired.

Amendment No. B3/4.5-19 '

f 1

RCIC Syst:m B 3/4.5.D B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.D. Reactor Core Isolation Cooling (RCIC) System '

BASES

BACKGROUND The RCIC is designed to provide makeup to the nuclear system as part of the planned operation for periods when the normal heat sink is 3

. unavailable. The Station Nuclear Safety Operational Analysis, FSAR Appendix G, shows that RCIC also serves as redundant makeup system on total loss of all offsite power in the event that HPCI is -

unavailable. In all other postulated accidents and transients, the ADS provides redundancy for the HPCI. -

SPECIFICATION The requirement that RCIC be operable when reactor coolant temperature is greater than 365 F is included in Specification 3.5.D.1 '

to clarify that RCIC need not be operable during certain testing (e.g.,

reactor vessel hydro testing at high reactor pressure and low reactor l coolant temperature). 365 F is approximately equal to the saturation

steam temperature at 150 psig.

ACTION Based on this and judgments on the reliability of the HPCI system, an allowable repair time of 14 days is specified.

SURVEILLANCES The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. To increase the availability of the core and containment cooling systems, the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested frequently. The pump and motor operated valves are tested in accordance with ASME B&PV Code,Section XI (lWP and IWV,

.except where specific relief is granted) to assure their operability.

The frequency and methods of Msting are described in the PNPS IST program. The PNPS IST Program is used to assess the operational readiness of pumps and valves that are safety-related or important to safety. When components are tested and found inoperable the impact on system operability is determined, and corrective action or Limiting Conditions of Operation are initiated. A simulated automatic actuation test once each cycle combined with code inservice testing of the pumps and valves is deemed to be adequate testing of these systems.

The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when required.

Amendment No. B3/4.5-20 l

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

. ADS SystIm B 3/4.5.E B 3/4.5 - CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.E. Automatic Depressurization (ADS) System BASES BACKGROUND This specification ensures the operability of the ADS under all conditions for which the automatic or manual depressurization of the r'uclear system is an essential response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low pressure coolant injection (LPCI) and the core spray systems

, can operate to protect the fuel barrier.

Because the Automatic Depressurization System does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further ,

conservatism to the CSCS. Performance analysis of the Automatic Depressurization System is considered only with respect to its depressurizing effect in conjunction with LPCI or Core Spray. There are four valves provided and each has a capacity of 800,000 lb/hr at a reactor pressure of 1125 psig.

APPLICABLE The limiting conditions for operating the ADS are derived from the

. SAFETY ANALYSIS Station Nuclear Safety Operational Analysis (FSAR Appendix G) and a detailed functional analysis of the ADS (FSAR Section 6).

ACTIONS The allowable out of service time for one ADS valve is determined as 14 days because of the redundancy and because of HPCI operability; therefore, redundant protection for the core with a small break in the

nuclear system is still available.

SURVEILLANCES The testing interval for the core and containment cooling systems is i based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. When components are tested and found inoperable the impact on system operability is determined, and corrective action or Limiting Conditions of Operation are initiated. A simulated automatic actuation test once each cycle combined with code inservice testing of the pumps and valves is deemed to be adequate testing of these systems. The ADS test circuit permits conti'nued surveillance on the operable relief valves to assure that they will be available if required.

The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when required.

' Amendment No. B3/4.5-21

i

!' *. Minimum Lcw Pressure Cceling cnd Diestl G:n:ratar Availability 3/4.5.F B 3/4,5 CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.F. _ Minimum Low Pressure Cooling and Diesel Generator Availability -

BASES BACKGROUND The purpose of Specification 3/4.5.F is to assure that adequate core cooling equipment is available at all times. If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only 2 LCPI pumps would be available.

It is during refueling outages that major maintenance is performed and during such time that all low pressure core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive components and recirculation system.

Specification 3.4.F.5 allows removal of ene CRD mechanism while the torus is in a drained condition without compromising core cooling capability. The available core cooling capability for a potential draining of the reactor vessel while this work is performed is based on an estimated drain rate of 300 gpm if the control rod blade sealis unseated. Flooding the refuel cavity and dryer / separator pool to elevation 114'-0" corresponds to approximately 350,000 gallons of water and will provide core cooling capaMlity in the event leakage trom the control rod drive does occur. A potential draining of the reactor vessel (via control rod blade leakage) would allow this water to enter into the toruc and after approximately 140,000 gallons have accumulated (needed to meet minimum NPSH requirements for the LPCI and/or core spray pumps), the torus would be able to serve as a common suction header. This would allow a closed loop operation of the LPCI system and the core spray system (once re-aligned) to the torus. In addition, the other core spray system is lined up to the condensate storage tanks which can supplement the refuel cavity and dryer / separator pool water to provide core flooding, if required.

Specification 3.9 must also be consulted to determine other requirements for the diesel generators.

3/4.5.G. Deleted Amendment No. B3/4.5-22

i

Maint:nanca cf Filled Discharge Pipe 3/4.5.H B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.H. Maintenance of Filled Discharge Pipe BASES BACKGROUND If the discharge piping of the core spray, LPCI system, HPCI, and RCIC are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. An analysis has been done which shows that if a water hammer were to occur at the time at which the system were required, the system would still perform its design function. However, to minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an operable condition.

SURVEILLANCE An acceptable method of ensuring that the lines are full is to vent at the high points. The monthly frequency is based on the gradual nature -

of void buildup in the ECCS piping, the procedural controls, and operating experience.

d A'mendment No. 83/4.5-23 l

ML20203D498

Text

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