Text

Proposed Tech Specs Section 3.5,converting to ITS
	ML18068A585
Person / Time
Site:	Palisades
Issue date:	05/03/1999
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18068A584	List: ML18068A583; ML18068A585;
References
	NUDOCS 9905100133
	Download: ML18068A585 (145)
	v • d • e

3.S.I 4.2

FSAR Sect1on REF ERE~~'

1. CONTROL ROOS 7.6.1.3

(;:) 2. CONTROL ROOS arthl Movement f all Rods 7.6.1.3 (M1nilllUlll of 6 In)

Pressur1zer (t4-)- 3. Safety Valves 4.3.7

4. M1tn Stea11

(~~

Set Point ~.3.4 Safety Valves

5. Refueling Syst Functioning 9.11.4 (i.~~ Interlocks 6, S1rvtc1 Wate Functioning 9.1.2 System Yalv Actu1t1on o (5'4 7.

SIS and RA Ev1lu1t1 4.7.l 3.4-

<9 t~~)

3,\ Verify proper Daily te11perature readings.

AmtndMnt No. H, G, m, ~. +ii, W, -tOe, 4-11 9905100133 990503 PDR ADOCK 05000255 p PDR

J .S. J

4. 6 .1

a. t ts shall bt perfonaed each reactor refuel i g interval.

A test fety injection signal w 11 be applied to init* te opent

n of the syste11. Tht s fety injection ind sh down cool 1ng syste pump motors may be dt*e ergized for this test. The system will t considered utishcto y if control board ind cHion and vis l observations indicate that all C04ftC)onents ha e rece1ved tne SI ty injection signal In ht proper sequence and timing (ie, tne 1 ropriatt p~ breakers all have opened and osed, and 111 alvts shall have c~l1t ~ their travel .

1. Syste11 test shall be perfon11d at each reactor refueling interva

.The test shall bt perfor"llltd th tht isolation valves In the s ay 1

/supply 1ints 1t the contain nt blocked closed. Operation of the systet1 is initiattd by tr ping the non11al actuation

j.

instruMntlt1on.

At least every ten ye s the spray nozzles open.

ied to bt

c. The test w111 considtrtd satisfactory if visual observations indicate all mponents have operated satisfacto ly.

4.6.3 ~

injection pu""'s, shut own cooling pumps, ind co tainment s shall be started at 1nterv1ls not to exceed t ree months Alternate 11anual star ing ~t,..tn control room onsole and the 1 cal breaker shall be pr ctictd 1n the test progra *

b. Ace table levels of ptrfo net shall bt that the pu ps start, rt h their rattd heads o recirculation flow, and o erate for at 1 st fifteen *1nutts.

4.6.4 val yu s~ 3s.1.1 **

Sfl..~.s.1.5 The Lo Pressure Saft y Injection f1 path shall bt erified OPERAS E within 7 d s prior to 11c reactor st1rtu by verifying

flow ontrol valve 4.24 open, and Its a1r su ply is isolated.

.AMndMnt Mo. 64-, ;.;, %, ~. 1-*, ~,

(\1' \ ~ 5 .V I 74 October 31. 1996

\

/ -h 5 df 5

-3.s.~

4.6 SAFETY INJECTIQH AHQ CQHTAIHt!ENT Spp.AY SYSTEMS TEST~

RP-..\

~.s.G-l

4. 6 .1 SR35.L5

.R.J.5"'.2.~

4.6.2

1. S1s e* test shill be perfon11 1t e1ch re1ctor refueling in rv1l.

T test*sh1ll be ptrfo,...d 1th the isol1t1on v1lves int spray pply lines 1t the cont11 nt blocKed closed. Oper1tio of the ystH is 1n1tf1ted by tri ping the nonul 1ctu1tion 1nstruMntlt1on.

At 1HSt shill be v open *

4.6.3 Tht test will bt c s1dtrtd s1t1sf1ctory if v1su1 observ1t1ons indic1t1 111 col!lp tnts 1.

JR ~.S. 2.*'-f sR.3s. c,. '1 b* Acee rue of trfol"'91nce shill b th1t the pumps st1rt, on rtcircul1t1on flow, 1nd oper1te for at

( 'S"~co 3 .I,,) lust 4.6.4 y1Jy11

a. y Injection T1nk f1 pith shill be verifi d OPERABLE days prior to 11ch r 1ctor st1rtup by vtr ying 11ch

( See)

~S./

1sol1t1on v1lvt is pen by observing v1l 1nd1c ion ind v11vt 1tstlf and locking open th 1ssoci1ted circ t breakers.

position b.

SR 3.5'.2.J

Fiee.o 4*24 AMndMnt Ho. ~' ~. 9', W, +;+, ~. 174 October 31, \ 996 J-c._

Of 5

ATTAC1'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

/

LA.1 CTS 4.6.4a states that SIT motor opera d isolation valve position is verified "by /

_II observing valve position indication a (the) valve itself." The method of how to .

verify SIT motor operated isolatio alve position is being removed since it is /

adequately covered in p.2,nt proce ures. This information provides detail of a pro9~ss which is not directly pertinent t the actual Surveillance Requirement, but rather/*

describes an acceptable metho of compliance. Since this detail is not necessai;#' to adequately describe the actu regulatory requirement, it can be moved to othf~

documents without an imp ct on safety. /

/

LA.2 CTS 4.6.4a requires a erification that the breakers for the SIT motor o~rated isolation valves are lo ked opened. Proposed SR 3.5.1.5 requires that.power f

be removed from the v ve operator but does not specify that the breakefmust be locked in the open positio . The intent of removing power from the valve pperator is to ensure that an act* e failure does not result in the undetected closq/e of an SIT valve.

Specifying that e breaker must be locked opened does not cons,ftute a requirement l

which must be et to ensure the assumptions of the safety analfsis are met, but simply provides am od for assuring the breaker remains open. T¥refore, placing this detail in la procedures is acceptable since it is not necess,ry to adequately describe the actu~l gulatory requirement and maintaining this infoymation in plant procedures 1 l

1 1

will not r sult in a significant impact on safety. This chap.'ge is consistent with I :

1 NURE -1432. ;*

LA.3 CTr{.2 Table 4.2.2 item 10 contains a surveillance 1equirement to verify SIT pressure j is ess than the high alarm. This requirement has ~~1en removed from the ITS and .1 aced in plant procedures. Neither the CTS (3.3.Ab) nor the ITS specify a high l ressure limit fot"'the SlTs since a maximum pr~sure is not an assumption in the safety Ii analysis. As such, it is not appropriate to co~in a surveillance requirement in the.ITS whose function is strictly to alert the operat<)f of an off normal condition when the I Operability of the affected component isfof iinpacted. Therefore, placing this detail in '

plant procedures is acceptable since it is ot 'necessary to adequately describe the actual ,

J

..____result in a significant impact on sat regulatory requirement and maintainin~Jthis information in plant procedures will not 11 j

Th,r..,, W'-r~ l'io 11 ftmouo.l of ~c.+o.1ls C.h~f'l~t.S ~rJ~oc.Jo.ftel wit-fl +111J Sfe;;.;~~on

(<Al ~.5.G-/ I Palisades Nuclear Plant Page 4 of 5 01/20/98 I- d

LESS RESTRICTIVE CHANGES (L)

ATTAt:HMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

L.1 CTS 3.3.2 requires the reactor to be placed in cold shutdown when one SIT is not restored to Operable status within the allowed outage time. Proposed ITS 3 .5 .1 Required Action C.l requires the plant to be placed in MODE 3 when one SIT is not restored to Operable status within the allowed outage time. Required Action C. l places the plant in a condition in which the SITs are no longer necessary to mitigate the consequences of an accident. In MODE 3 and below, the rate of PCS blowdown during a LOCA is such that the HPSI and LPSI pumps can provide adequate injection to ensure peak cladding temperature remains below the 10 CFR 50.46 limit of 2200°F.

Only requiring the plant to be placed in a mode in which the LCO does not apply is consistent with the philosophy of NUREG-1432.

L.2 A new Condition, Required Action and Completion Time is proposed for the case where one SIT is inoperable due to its boron concentration not being within limits.

This addition in the ITS allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore boron versus the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the CTS.

The SITs are passive devices and the boron concentration is relatively stable.

Therefore, deviations from the normal boron concentration band are generally small

and the impact on a LOCA is minor. This is because boiling of the emergency core cooling water during the reflood phase of a LOCA concentrates the boron in the saturated liquid that remains in the core. In addition, the volume of the SIT is still available for injection. Since the boron requirements are based on the average boron concentration of the total volume of three SITs, the consequences are less severe than they would be if an SIT were not available for injection.

L.3

Palisades Nuclear Plant Page 5 of 5 01/20/98

/-e.

INSERT- L.4 CTS 4.6.4a and CTS 4.2, Table 4.2.2 item #10 contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion. Specifically, CTS 4.6.4a states that SIT motor operated isolation valve position is verified "by observing valve position indication and (the) valve itself" In addition, CTS 4.6.4a also requires a verification that the breakers for the SIT motor operated isolation valves are locked opened. The intent of verifying valve position and removing power from the valve operator is to ensure that the valve is in its correct position and that an active failure does not result in the undetected closure of a valve. Specifying that valve position be verified "by observing valve position indication and (the) valve itself', or stating that the "breaker must be locked opened" do not constitute requirements assumed in the safety analyses. Rather, they simply provide a method for assuring valve and breaker position. CTS 4.2 Table 4.2.2 item 10 contains a surveillance requirement to verify SIT pressure is less than the high alarm. The purpose of this alarm is to alert the operator of an off normal condition (e.g., nitrogen inleakage or check valve back leakage). This requirement has been deleted since the safety analyses do not

. assume a maximum SIT pressure. Thus, it is not appropriate to contain a requirement whose function is strictly to alert the operator of an off normal condition when the Operability of the affected component is not impacted. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432 .

)-f

ATTAC1'ENT. 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.2, ECCS - OPERATING LA.1 (continued)

Removing the details of the ECCS from the CTS and placing them in the Bases of the ITS is acceptable since these details are not pertinent to the actual requirements.

Placing these details in the Bases provides adequate assurance that they will be maintained since the Bases are controlled by the Bases Control Program proposed in ITS Chapter 5.0. This change is consistent with NUREG-1432.

LA.2 CTS 4.6.3a states th "alternate manual starting (of the safety injection pum sand shutdown cooling a mps) between the control room console and the local b eaker shall be practiced in th test program." The ability to demonstrate the manual arting capability of the CCS pumps from various plant locations is not assum in the safety analyses and is ot relevant to demonstrating that the pumps are capabi "ty of meeting their intended afety function. As such, these details have been remoVJ d from the CTS and placed i plant procedures. Since these details are not necessary o describe actual regulatory quirements, they can be moved to licensee control wi ut a significant impact on afety. This change is consistent with NUREG-1432.

~

LA.3 CTS 4. . lA contains the requirement for testing the Safety Inje tion System. The CTS states at satisfactory results of the test are indicated by "con ol board indication and visu observation." In the ITS, this level of detail is not in ded in the Surveillance Re irements since it is not pertinent to the actual require nt. The intent if this test is o verify that components which receive an actuation si al actuate to their correct sition. The method of verification (i.e., control board ndication or visual observation) is more appropriately stated in plant proce ures. As such, these details have been moved to plant procedure. Since these de s are not necessary to describe actual regulatory reqwrements:they can be moved t licensee control without a significant impact on safety. This change is consis nt with NUREG-1432 .

Palisades Nuclear Plant Page 6 of 7 01/20/98

ATTACIThfENT 3 DISCUSSION OF .CHANGES SPECIFICATION 3.5.2, ECCS - OPERATING LESS RESTRICTIVE CHANGES (L)

L.1 CTS 3. 3. 2 requires the reactor to be placed in cold shutdown within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months when an ECCS component is not restored to Operable status within the allowed outage time.

Proposed ITS 3.5.2 Required Action B.2 requires the PCS temperature to be reduced to

< 325 °F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when an ECCS train is not restored to Operable status within the allowed outage time. Required Action B.2 places the plant in a condition in which two trains of ECCS are no longer necessary to mitigate the consequences of an accident assuming a single failure. In MODE 3 with PCS temperature < 325°F, ECCS operational requirements are relaxed due to stable plant conditions and the reduced probability of a DBA. ECCS requirements in MODE 3 with PCS temperature

< 325°F are addressed by proposed ITS 3.5.3, ECCS - Shutdown. Requiring the PiAI plant to be placed in a mode in which the LCO no longer applies is consistent with the ~,5'. ~-I philosophy of NUREG-1432. CJJ'lt,,. I C.1S 3.3.Z...b C.TS 3.1:1.2c., c~ 3J.2d, a.rJ C.75 3.3-Z:f tt~u1r.:. thc.ir ~ftc.+i.Ut tf\O/tJ'ablt. cCCl ~f L.2 c; CTS 3.3.;i!e reeiaires taat aa iaspeFat;Jl8 WJ>£1 pYIRp be restored to Operable status ~

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Proposed ITS Condition A has revised the CTS to allow one or more ECCS train to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least 100 % of the ECCS flow

equivalent to a single Operable ECCS train is available. Thus, the ITS allows aa IIPSI subsystem, or a combination of HPSI and LPSI subsystem~to be inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the flow assumed to be delivered by a single ECCS train is available. By stipulating a 100% ECCS flow equivalent, proposed Condition A preserves the safety function of the ECCS system while allowing some period of time x

for correcting ECCS component inoperabilities. This allowance is acceptable because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. This change is supported by reliability analyses discussed in an NRC Memorandum to V. Stello, Jr., from R.L.

Baer, "Recommended Interim Revisions to LCOs for ECCS Components,"

December 1, 1975. This change is consistent with NUREG-1432.

L3 l\Jt.uJ - St~ l!"Sc.r+

s <.-C..

I..r-JSl~l f?AI 3. S. G-/

L.Lf {'lt-c...J

Palisades Nuclear Plant Page 7of7 01/20/98 1-h

INSERT -DOC 3.5.2 - L.4 CTS 4.6: la and CTS 4.6.3a contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion.

Specifically, CTS 4.6. IA contains the requirement for testing the Safety Injection System. The CTS states that satisfactory results of the test are indicated by "control board indication and visual observation." The intent of this test is to verify that components which receive an actuation signal actuate to their correct position. Specifying that valve position be verified "by "control board indication and visual observation." does not constitute a requirement assumed in the safety analyses. Rather, it simply provide a method for assuring valve position. CTS 4.6.3a states that "alternate manual starting (of the safety injection pumps and shutdown cooling pumps) between the control room console and the local breaker shall be practiced in the test program." The ability to demonstrate the manual starting capability of the ECCS pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capability of meeting their intended safety function. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432 .

J *-/

ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

1. (continued)

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed mitigatory function of the remaining Operable SITs or other emergency core cooling system structures or components. Thus, the consequence of an accident occurring during the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowed outage time presently specified in the technical specifications is the same as the consequences for an accident occurring during an allowed outage time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made.to plant parameters which govern normal plant operation. The proposed change will continue to ensure prompt restoration of an inoperable SIT to re-establish compliance with the limiting condition for operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change increases the allowed outage time from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for an SIT whose boron concentration is not within limits, or when level or pressure of an SIT can not be verified. The margin of safety afforded by the SITs is in their ability to supply water to the reactor vessel during a LOCA and thus help ensure that the acceptance criteria, established by 10 CFR 50.46 for the emergency core cooling system, will be met. Since the SITs are passive components, their boron concentration, level and pressure are relatively stable and deviations in these parameters are generally small. Thus, the impact on a LOCA as a result of one of these parameters being outside their limit is minor since the contents of the inoperable SIT will still be available for injection in the event of a LOCA. In addition, an extension in the allowed outage time provides some period of time to restore an inoperable SIT to Operable status without initiating a plant shutdown which increases the chance of an undesired plant transient. As such, any reduction in a margin of safety resulting from an extended allowed outage time would most likely be offset from the benefits gained from restoring the inoperable SIT and precluding an unnecessary plant shutdown .

Therefore, this change does not involve a significant reduction in a margin of safety.

L-3 Ne.w RA1 Ei.S.\*2- Cot'<\~ ~l. x

( Palisades Nuclear Plant Page 3of3 01/20/98 L.L/ Ntw RA-I 3.5.G-I 1-j r

INSERT - NSHC 3.5. 1 - L.4 CTS 4.6.4a and CTS 4.2, Table 4.2.2 item #10 contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion. Specifically, CTS 4.6.4a states that SIT motor operated isolation valve position is verified "by observing valve position indication and (the) valve itself" In addition, CTS 4.6.4a also requires a verification that the breakers for the SIT motor operated isolation valves are locked opened. The intent of verifying valve position and removing power from the valve operator is to ensure that the valve is in its correct position and that an active failure does not result in the undetected closure of a valve. Specifying that valve position be verified "by observing valve position indication and (the) valve itself', or stating that the "breaker must be locked opened" do not constitute requirements assumed in the safety analyses. Rather, they simply provide a method for assuring valve and breaker position. CTS 4.2 Table 4.2.2 item 10 contains a surveillance requirement to verify SIT pressure is less than the high alarm. The purpose of this alarm is to alert the operator of an off normal condition (e.g., nitrogen inleakage or check valve back leakage). This requirement has been deleted since the safety analyses do not assume a maximum SIT pressure. Thus, it is not appropriate to contain a requirement whose function is strictly to alert the operator of an off normal condition when the Operability of the affected component is not impacted. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event.

The proposed change deletes details from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated .

/-A

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

The proposed change deletes detail from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes deletes details from the Technical Specifications. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

1-L

ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING

" ~Al 3i.z. .I

3. Does this change involve a significant reduction in a margin of safety? °""°' 1

a. n 1noll!.i1llilc. Ceo The proposed change extends the allowed outage time for a ~ IIPSI or ~ump Clin-T (

from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and the combination of an HPSI ~ and LPSI 1

from {

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provide at least 100% flow equivalent to a single ECCS train is available. The function of the ECCS is to provide core cooling and negative reactivity to ensure the core is protected following various accidents. By stipulating a 100%

ECCS flow equivalent as a condition to extend the inoperability of an ECCS train, the safety function of the ECCS system is preserved. This is because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. As such, the margin of safety associated with the extended allowed outage time remains unchanged. Therefore, this change does not involve a significant reduction in a margin of safety.

P-.111 3. S. l L.3 Ne..uJ C...ort\M "° ]_

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RAI ?J.5.b-1 Se..c... DJ)lfl1 x Palisades Nuclear Plant Page 4 of 4 01120/98 1-fY)

INSERT - NSHC 3.5.2 - L.4 CTS 4.6. la and CTS 4.6.3a contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion.

Specifically, CTS 4.6. IA contains the requirement for testing the Safety Injection System. The CTS states that satisfactory results of the test are indicated by "control board indication and visual observation." The intent if this test is to verify that components which receive an actuation signal actuate to their correct position. Specifying that valve position be verified "by "control board indication and visual observation." does not constitute a requirement assumed in the safety analyses. Rather, it simply provide a method for assuring valve position. CTS 4.6.3a states that "alternate manual starting (of the safety injection pumps and shutdown cooling pumps) between the control room console and the local breaker shall be practiced in the test program." The ability to demonstrate the manual starting capability of the ECCS pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capability of meeting their intended safety function. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event.

The proposed change deletes details from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

I -"7_

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

The proposed change deletes detail from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes deletes details from the Technical Specifications. Removal of these*

details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety .

/ -()

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION

NRC REQUEST:

SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS 3.5.1, Safety Injections Tanks (SITs) 3.5.1-1 CTS 3.3.1 ITS 3.5.1 Applicability and Required Actions C.1 and C.2 DOC A.2, L.1 and JFD #5 CTS 3.3.1 states that "The reactor shall not be made critical ... unless all of the following conditions are met:" Item b is applicable for the Safety Injection Tanks. ITS 3.5.1 Applicability is proposed to be Mode 1 and 2. STS 3.5.1 Applicability is Mode 1 and 2, and Mode 3 with RCS pressure greater than 700 psia.

Comment: JFD #5 explains that the SITs are pressurized to greater than 200 psig and filled using the HPSI pumps and injection lines and that attempting to fill the SITs with PCS pressure below the HPSI shutoff head would result in flow opening the loop check valves and entering the PCS. The JFD also explains how the parameters of the SITs are currently verified within limits after plant heatup and before the approach to criticality.

The Applicability of SITs should be based upon when they are needed to mitigate the assumed accident when the PCS is at elevated temperatures and pressures .

Therefore, shouldn't SITs have a similar Applicability to the ECCS trains in ITS 3.5.2 which begins in Mode 3 after PCS reaches 325°F or an Applicability based on when PCS pressure exceeds the SIT cover pressure of 200 psig? The resolution of the Applicability issue may require the addition of Required Action C.2 to "Reduce PCS temperature to < 325°F" or "Reduce PCS pressure to < 200 psig" with a technically justified Completion Time .

(continued) 2

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS .

RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION

NRC REQUEST:

3.5.1-1 (continued)

SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS Consumers Energy Response:

The Applicability for the SITs in the ITS is stated as Modes 1 and 2 consistent with the requirements of the CTS. The SITs are credited in both the large and small break LOCA analyses which assume an initial reactor power of 100% RTP. No specific analysis has been performed to quantify the contribution from the SITs for events initiated at hot zero power. Similarly, no specific issue has been identified which challenges the basis for the existing CTS requirement. In Mode 3, additional thermal margin is gained from the peak clad temperature determined in the LOCA analyses due to the significant reduction in the thermal energy stored in the fuel. In addition, an increase in thermal margin is realized from the reduction in primary coolant sensible heat due to a cooldown from PCS average temperature 555°F in MODE 1 to 532°F or lower in MODE 3. Although some adiabatic heatup is expected to occur as a result of a LOCA in MODE 3, the resultant decrease in PCS blowdown rate from the reduction in stored thermal energy is such that the ECCS pumps (two HPSI and two LPSI) can provide adequate core injection, without contribution from the SITs, to ensure peak clad

temperature remains below 2200°F. As such, the Applicability as stated in the ITS represents the plant conditions at which the SITs are relied upon to mitigate the consequences of an accident.

Affected Submittal Pages:

No page changes .

3

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION

NRC REQUEST:

3.5.1-2 SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS CTS 3.3.2, CTS 3.3.2 e and f ITS 3.5.1 Action A, B, C, and D DOC A.5 CTS 3.3.2 states that During power operation the requirements of (CTS) 3.3.1 may 11 be modified to allow one of the following conditions to be true at any one time 11 ITS 3.5.1 has converted these requirements into the improved STS format.

Comment #1: DOC A.5 appears to be composed of several change issues and should be re-written to separate these issues.

DOC A.5 appears to apply to Action B for justification of the one hour Completion Time. The one hour is retained which is acceptable and it is an administrative change as explained in the 7th through 10th sentences. The 11th sentence discusses a relaxation of the Completion Time to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months which is not identified in the ITS, nor is it discussed in DOC L.1, as noted in this DOC A.5. This 11th sentence appears to be extraneous and should be deleted.

Consumers Energy Response:

As a result of our response to NRC Request 3.5.1-2, Comment #2, DOC A.5 has been deleted in its entirety. Thus, the extraneous information in the 11th sentence of that DOC has also been deleted.

Affected Submittal Pages:

See response to RAI 3.5.1-2 comment #2 .

4

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.1-2 CTS 3.3.2, CTS 3.3.2 e and f ITS 3.5.1 Action A, B, C, and D DOC A.5 CTS 3.3.2 states that "During power operation the requirements of (CTS) 3.3.1 may be modified to allow one of the following conditions to be true at any one time".

ITS 3.5.1 has converted these requirements into the improved STS format.

Comment #2: DOC A.5 appears to be composed of several change issues and should be re-written to separate these issues.

The fifth and sixth sentences explain how the reformatting of these CTS requirements have resulted in a 11 generic 11 less restrictive technical change for several ITS 3.5 LCOs. The concurrent inoperability permitted by separate ITS LCOs would invoke a CTS LCO 3.0.3 plant shutdown under the CTS 3.3.2 requirements.

This is a less restrictive change which should be separately identified.

Consumers Energy Response:

A new "less restrictive change" (DOC L.3) has been provided in ITS 3.5.1 and ITS 3.5.2 to address the allowance for concurrent ECCS component inoperabilities precluded by CTS 3.3.2. For ITS 3.5.1, this change supersedes the justification previously provided in DOC A.5.

Affected Submittal Pages:

Att 3, CTS pg 3-29a, (ITS 3.5.1, 2 of 5)

Att 3, CTS pg 3-29a, (ITS 3.5.2, 2 of 5)

Att 3, DOC 3.5.1, pg 2 of 5 Att 3, DOC 3.5.1, pg 5 of 5 Att 3, DOC 3.5.2, pg 7 of 7 Att 4, NSHC 3.5.1, pg 3 of 3 Att 4, NSHC 3.5.2, pg 4 of 4

5

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3*291 A.mtnament No. i+, 5-t, 172 September 26, l 996

..5 -('.L, IL of 5

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ithin tht

0

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tht pu~p is rtstortd to op1r1blt status within hours. ?Z. ---...

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d.

l*2h Amendment No. c-f., 9-t, 17Z September 26, 1996

.~-b

ATTACHMENT 3

A.4 DISCUSSION OJ:(' CHANGES SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

CTS 3. 3 .1 has been revised to include a new Condition, Required Action and Completion Time to address two or more inoperable SITs. The proposed change requires an immediate entry into LCO 3.0.3. This proposed change does not impose any new requirement on plant operation since the CTS currently requires entry into LCO 3.0.3 with two or more inoperable SITs. The intent of this change is to eliminate the potential for the misinterpretation of the Required Actions if two or more SITs were inoperable for different reasons (i.e., boron concentration not within limits in one SIT concurrent with the inability to verify pressure or level in another SIT). Since this change only impacts the presentation of the CTS requirements, it is considered administrative in nature.

**--~-~"~*~-----:r-----------r--~

A.5 The purpose of this change is t discuss how the requirements for the SITs re unchanged based strictly on difference in format of the ISTS and propose x

from the CTS. CTS 3.3.2 s tes that "during power operation the requirem ts of (CTS) 3. 3 .1 may be modifi Cl to allow one of the following conditions to b true at any one time." Contained int e list of conditions for CTS 3.3.2 are one Safi Injection Tank (SIT) for a period one hour, one Low Pressure Safety Injection PSI) pump for a period of 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , one High Pressure Safety Injection (HPSI) pu p for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , one shut wn heat exchanger for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a d associated system (i.e., LPSI or PSI) valves for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In the STS and proposed ITS, the SITs are a ressed in a separate specification from other CS related components (LPSI nd HPSI pumps, shutdown heat exchanger an related system valves). Therefo e, the ISTS and proposed ITS allow the inoper bility of an SIT to exist concurrent with other inoperable ECCS components. If *s same concurrent inoperability o curred in the CTS, then the actions of LCO 3 .. 3 would be invoked.

CTS LCO 3. . 3 would allow one hour to initiate actions to ace the plant in a condition in hich the specification does not apply and six ours to place the plant in Hot Stand . As a result of reformatting the CTS, the IS S would also allow one hour to restore an inoperable SIT to an operable status or, re ire the plant to be placed in Mode 3 ith 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . As such, the requirements of th CTS (which invoked LCO 3 0.3) would be equivalent to the requirements f the ISTS. Thus, this change in form is characterized as administrative in nature. lthough the actual Completion T" for an inoperable SIT has been extended fro one hour as allowed in the CTS, to 24 ours as specified in the ITS, this change is c racterized as Less Restrictive and is d scribed inde endentl in Discussion of Chan L. l .

ll.5 NGW

"- ~A I ~ .S. l *2- Comm~ 3 Palisades Nuclear Plant Page 2 of 5 01/20/98

ATTAf:HMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.1, SAFETY INJECTION TAi"1'KS (SITs)

LESS RESTRICTIVE CHANGES (L)

L.1 CTS 3.3.2 requires the reactor to be placed in cold shutdown when one SIT is not restored to Operable status within the allowed outage time. Proposed ITS 3 .5 .1 Required Action C. l requires the plant to be placed in MODE 3 when one SIT is not restored to Operable status within the allowed outage time. Required A~tion C. l places the plant in a condition in which the SITs are no longer necessary to mitigate the consequences of an accident. In MODE 3 and below, the rate of PCS blowdown during a LOCA is such that the HPSI and LPSI pumps can provide adequate injection to ensure peak cladding temperature remains below the 10 CFR 50.46 limit of 2200°F.

Only requiring the plant to be placed in a mode in which the LCO does not apply is

consistent with the philosophy of NUREG-1432.

L.2 A new Condition, Required Action and Completion Time is proposed for the case where one SIT is inoperable due to its boron concentration not being within limits.

This addition in the ITS allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore boron versus the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the CTS.

The SITs are passive devices and the boron concentration is relatively stable.

Therefore, deviations from the normal boron concentration band are generally small and the impact on a LOCA is minor. This is because boiling of the emergency core cooling water during the reflood phase of a LOCA concentrates the boron in the saturated liquid that remains in the core. In addition, the volume of the SIT is still available for injection. Since the boron requirements are based on the average boron concentration of the total volume of three SITs, the consequences are less severe than they would be if an SIT were not available for injection .

Palisades Nuclear Plant Page 5 of 5 01/20/98 5-d

INSERT - DOC 3.5. 1 - L.3 CTS 3.3.2 contains a provision which allows one of the ECCS components required by CTS 3 .3 .1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements ofLCO 3.0.3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss ofECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a"condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss of ECCS function will not occur. This is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432 .

5-e

ATT AC!Th-IENT 3 DISCUSSION OF CHANGES

LESS RESTRICTIVE CHANGES (L)

L.1 SPECIFICATION 3.5.2, ECCS - OPERATING CTS 3.3.2 requires the reactor to be placed in cold shutdown within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months when an ECCS component is not restored to Operable status within the allowed outage time.

Proposed ITS 3.5.2 Required Action B.2 requires the PCS temperature to be reduced to

< 325°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when an ECCS train is not restored to Operable status within the allowed outage time. Required Action B.2 places the plant in a condition in which two trains of ECCS are no longer necessary to mitigate the consequences of an accident assuming a single failure. In MODE 3 with PCS temperature < 325°F, ECCS operational requirements are relaxed due to stable plant conditions and the reduced probability of a DBA. ECCS requirements in MODE 3 with PCS temperature ,

< 325°F are addressed by proposed ITS 3.5.3, ECCS - Shutdown. Requiring the RA/

plant to be placed in a mode in which the LCO no longer applies is consistent with the ~.5'* ~*I philosophy of NUREG-1432. Co/'lf11t I C.1S 3.3-Lb C.TS3.~.2', C1SJJ2d.a.rJ C7S 3.3.U" rt"tu'~ +nc.ir ~fe"ti*uc 1f\oft.ro.bk cCd ~f L. 2 G CTS 3. 3~e reeiliire& tl:lat aA iAepera9li WP~I pYR:ifl be restored to Operable status 'f...

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Proposed ITS Condition A has revised the CTS to allow one or more ECCS train to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least 100 % of the ECCS flow equivalent to a single Operable ECCS train is available. Thus, the ITS allows aa IIPSI X imbsystem, or a combination of HPSI and LPSI subsystem~to be inoperable for up to 'I. "i-72 hours provided the flow assumed to be delivered by a single ECCS train is available. By stipulating a 100% ECCS flow equivalent, proposed Condition A preserves the safety function of the ECCS system while allowing some period of time for correcting ECCS component inoperabilities. This allowance is acceptable because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. This change is supported by reliability analyses discussed in an NRC Memorandum to V. Stello, Jr., from R.L.

Baer, "Reconunended Interim Revisions to LCOs for ECCS Components,"

December 1, 1975. This change is consistent with NUREG-1432.

L3 f?AI 3.s.G-/

Palisades Nuclear Plant Page 7 of 7 01120/98 5-f

INSERT-DOC 3.5.2 - L.3 CTS 3.3.2 contains a provision which allows one of the ECCS components required by CTS 3. 3 .1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements of LCO 3. 0. 3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss ofECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a"condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss ofECCS function will not occur. This is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432.

j_a/

ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

1. (continued)

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed mitigatory function of the remaining Operable SITs or other emergency core cooling system structures or components. Thus, the consequence of an accident occurring during the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowed outage time presently specified in the technical specifications is the same as the consequences for an accident occurring during an allowed outage time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluatep..

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to ensure prompt restoration of an inoperable SIT to re-establish compliance with the limiting condition for operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change increases the allowed outage time from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for an SIT whose boron concentration is not within limits, or when level or pressure of an SIT can not be verified. The margin of safety afforded by the SITs is in their ability to supply water to the reactor vessel during a LOCA and thus help ensure that the acceptance criteria, established by 10 CFR 50.46 for the emergency core cooling system, will be met. Since the SITs are passive components, their boron concentration, level and pressure are relatively stable and deviations in these parameters are generally small. Thus, the impact on a LOCA as a result of one of these parameters being outside their limit is minor since the contents of the inoperable SIT will still be available for injection in the event of a LOCA. In addition, an extension in the allowed outage time provides some period of time to restore an inoperable SIT to Operable status without initiating a plant shutdown which increases the chance of an undesired plant transient. As such, any reduction in a margin of safety resulting from an extended allowed outage time would most likely be offset from the benefits gained from restoring the inoperable SIT and precluding an unnecessary plant shutdown.

Therefore, this change does not involve a significant reduction in a margin of safety.

L,3 Ntw RA1 ~.S,\ Co,....~ ~z_

x

( Palisades Nuclear Plant Page 3 of 3 01/20/98 L.l/ N<.w f<.Prl 3.S.G-1 r 5-~

INSERT - NSHC 3.5. 1 - L.3 CTS 3.3.2 contains a provision which allows one of the ECCS components required by CTS 3.3.1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements ofLCO 3.0.3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss ofECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a"condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss ofECCS function will not occur. This is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change will allow multiple components associated with the Emergency Core Cooling System (ECCS) to be simulataneously inoperable for a specified time provided the core cooling function is not lost. An extension in the allowed outage time for inoperable components is not assumed to be an initiator of any evaluated accident. Therefore, extending the allowed outage time for multiple component inoperabilities while preserving the overall function of the specified requirement does not involve a significant increase in the probability of an accident previously evaluated.

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change only extends the time multiple ECCS components can be made inoperable without requiring a plant shutdown provided the functional capability of the system is maintained. Thus the consequences of an accident occurring during the allowed outage time presently specified in the technical specifications is the same as the consequences for an accident occurring during the proposed allowed outage time in the ITS.

Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to assure the approprite ECCS components are Operable to fulfill the core cooling function during a DBA.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change increases the time mutiple ECCS components can be made inoperable provided the core cooling function is maintained. The margin of safety afforded by the ECCS is related to the assurance that the acceptance criteria setforth in 10 CPR 50.46 will be met following a LOCA. Acceptable consequences following a LOCA can be achieved by a single train of ECCS components and 3 of 4 Safety Injection Tanks. The proposed change continues to ensure at least 100% of the ECCS flow equivalent to a single Operable ECCS train, and 3 of 4 Safety Injection Tanks, are available. As such, removing more than one ECCS component at any given time does not affect affect the capability of the ECCS to perform its intended function. Any

reduction in a margin of safety resulting from an extended allowed outage time would mostly likely be offset from the benefits gained from restoring* the inoperable components to operation and thereby precluding an unnecessary shutdown. Therefore, the proposed change does not involve a significant reduction in a margin of safety .

5-j

ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERA TING

. ~Al 3iE5 z.-1 S"-1

3. Does this change involve a significant reduction in a margin of safety? °"'rn 1 a.n 1no1r.r>ihlo CCC5 (

The proposed change extends the allowed outage time for a - MPSI or ~l:lffl~

from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and the combination of an HPSI ,, and LPSI ' from {

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provide at least 1003 flow equivalent to a single ECCS train is available. The function of the ECCS is to provide core cooling and negative reactivity to ensure the core is protected following various accidents. By stipulating a 100 3 ECCS flow equivalent as a condition to extend the inoperability of an ECCS train, the safety function of the ECCS system is preserved. This is because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. As such, the margin of safety associated with the extended allowed outage time remains unchanged. Therefore, this change does not involve a significant reduction in a margin of safety.

r\AI 3.5.1-2..

L.3 f\k.uJ C...om IY'I ~ 1-See TuSc:RT L.lf (Je.~

RAI 3.5.b-/

Se..~ :r;,J'\U..1 x

Palisades Nuclear Plant Page 4 of 4 01/20/98

INSERT- NSHC 3.5.2 - L.3 CTS 3.3.2 contains a provision which allows one of the ECCS components required by CTS 3.3.1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements ofLCO 3.0.3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss ofECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a"condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss of ECCS function will not occur. This is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change will allow multiple components associated with the Emergency Core Cooling System (ECCS) to be simulataneously inoperable for a specified time provided the core cooling function is not lost. An extension in the allowed outage time for inoperable components is not assumed to be an initiator of any evaluated accident. Therefore, extending the allowed outage time for multiple component inoperabilities while preserving the overall function of the specified requirement does not involve a significant increase in the probability of an accident previously evaluated.

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change only extends the time multiple ECCS components can be made inoperable without requiring a plant shutdown provided the functional capability of the system is maintained. Thus the consequences of an accident occurring during the allowed outage time presently specified in the technical specifications is the same as the consequences for an accident occurring during the proposed allowed outage _time in the ITS .

Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

5-l

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to assure the approprite ECCS components are Operable to fulfill the core cooling function during a DBA.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change increases the time mutiple ECCS components can be made inoperable provided the core cooling function is maintained. The margin of safety afforded by the ECCS is related to the assurance that the acceptance criteria setforth in 10 CPR 50.46 will be met following a LOCA. Acceptable consequences following a LOCA can be achieved by a single train of ECCS components and 3 of 4 Safety Injection Tanks. The proposed change continues to ensure at least 1003 of the ECCS flow equivalent to a single Operable ECCS train, and 3 of 4 Safety Injection Tanks, are available. As such, removing more than one ECCS component at any given time does not affect affect the capability of the ECCS to perform its intended function. Any reduction in a margin of safety resulting from an extended allowed outage time would mostly likely be offset from the benefits gained from restoring the inoperable components to operation and thereby precluding an unnecessary shutdown. Therefore, the proposed change does not involve a significant reduction in a margin of safety .

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.1-2 CTS 3.3.2, CTS 3.3.2 e and f ITS 3.5.1 Action A, B, C, and D DOC A.5 CTS 3.3.2 states that During power operation the requirements of (CTS) 3.3.1 11 may be modified to allow one of the following conditions to be true at any one time 11

ITS 3.5.1 has converted these requirements into the improved STS format.

Comment #3: DOC A.5 appears to be composed of several change issues and should be re-written to separate these issues.

CTS 3.3.2 e and f have been removed per the CTS markup as only applicable for ITS 3.5.2. It appears that these two CTS requirements would still be applicable to the SITs under ITS 3.5.1. Provide a new DOC for these CTS requirements to explain how they are converted into the ITS 3.5.1 requirements.

Consumers Energy Response:

The markup of CTS page 3-29a (for ITS 3.5.1) and DOC A.5 have been revised to explain how the requirements of CTS 3.3.2e and CTS 3.3.2f have been converted i n ITS 3 . 5 . 1.

Affected Submittal Pages:

Att 3, CTS pg 3-29a (ITS 3.5.1, pg 2 of 5)

Att 3, DOC 3.5.1, pg 2 of 5

6

.J.3

3. 3' 2 EMERGENCY COSE CQQLING sySIQ1 (Cont'd) x C.01.1i) c.

§

(]

c:v...,n & i. One s1fety injection tink **Y bt inoptrible for 1 period of no mor*J~~'

thin one hour. ~

rb. p~ lliY bt inoper1ble stitus w1thln 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G~~) IC. 9h*pressure Sifety lnjejt1on p~ miy bt inoper1b is re5tored to oceribl1 st1tus w1thin 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

f. . Any Vi~l* interlock orliti pp 1ssociited with tht s1f ty injection ind shut own cooling systt a~d which is not covert under 3.3.2*

ibovt b , which is rtqui to function durin9 1c id1nt condit ns, ..1 bt 1nop1 pl1 for

period of no rt thin R/11 3. S. (.. z Corr."'~"r .+ 3

\

< Ao D C.0..iD. A) ([fJ

(ADI> ~tJ.D D) 8

3*29&

Amendment Ho . ..+, s.t-, 172 Sept1mbtr 26. 1996 0-o.- ;<_ af 5

ATTACHMENT 3 DISCUSSION O.tt' CHANGES

A.4 SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

CTS 3. 3 .1 has been revised to include a new Condition, Required Action and Completion Time to address two or more inoperable SITs. The proposed change requires an immediate entry into LCO 3.0.3. This proposed change does not impose any new requirement on plant operation since the CTS currently requires entry into LCO 3.0.3 with two or more inoperable SITs. The intent of this change is to eliminate the potential for the misinterpretation of the Required Actions if two or more SITs were inoperable for different reasons (i.e., boron concentration not within limits in one SIT concurrent with the inability to verify pressure or level in another SIT). Since this change only impacts the presentation of the CTS requirements, it is considered administrative in nature.

A.5 The purpose of this change is t discuss how the requirements for the SITs re unchanged based strictly on t difference in format of the ISTS and propose x

from the CTS. CTS 3. 3. 2 st tes that "during power operation the requirem ts of (CTS) 3. 3 .1 may be modifi a to allow one of the following conditions to b true at any one time." Contained in t e list of conditions for CTS 3 .3 .2 are one SaD Injection Tanlc (SIT) for a period one hour, one Low Pressure Safety Injection PSI) pump for a period of 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , one High Pressure Safety Injection (HPSI) pu p for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , one shut wn heat exchanger for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a d associated system (i.e., LPSI or PSI) valves for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In the TS and proposed ITS, the SITs are a ressed in a separate specification from other CS related components (LPSI nd HPSI pumps, shutdown heat exchanger an related system valves). Therefo e, the ISTS and proposed ITS allow the inoper bility of an SIT to exist concurrent with other inoperable ECCS components. If his same concurrent inoperability o curred in the CTS, then the actions of LCO 3 .. 3 would be invoked.

CTS LCO 3. . 3 would allow one hour to initiate actions to ace the plant in a condition in hich the specification does not apply and six ours to place the plant in Hot Stand . As a result of reformatting the CTS, the IS S would also allow one hour to restore an inoperable SIT to an operable status or, re ire the plant to be placed in Mode 3 ith 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . As such, the requirements of th CTS (which invoked LCO 3 0.3) would be equivalent to the requirements f the ISTS. Thus, this change in form is characterized as administrative in nature. !though the actual Completion Ti for an inoperable SIT has been extended fro one hour as allowed in the CTS, to 24 ours as specified in the ITS, this change is c racterized as Less Restrictive and is d scribed inde endentl in Discussion of Chan L.1.

A.5 rJe..w

~ f<A I 3,5. l *2.. C.ornm 1t 3 -

Palisades Nuclear Plant Page 2 of 5 01/20/98

INSERT - DOC 3.5. l - A.5 CTS 3.3 .2e provides the required actions for any valves, interlocks or piping directly associated with the SITs. CTS 3.3.2f provides the required actions for any valves, interlocks or piping associated with the SITs which is not covered by CTS 3.3.2e. In the ITS, these same valves, interlocks and pipes are addressed by the definition of Operability. An earlier version of the CTS defined Operable as "a system or component is operable if it is capable of fulfilling its design function." As such, to ensure all the components necessary to fulfill the safety function of a system, subsystem or train were adequately covered, the CTS provided two distinct required actions. Subsequent changes to the CTS have redefined the term "Operable" such that it is no longer necessary to provide two different required actions. Since the revised definition of Operable in the CTS is consistent with the definition of Operable in the ITS, the required actions ofCTS 3.3.2e and CTS 3.3.2f are no longer necessary.

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.1-3 CTS 4.2, Table 4.2.2, Item #10 ITS SR 3. 5 .1. 3 DOC LA.3 and JFD #7 CTS Table 4.2.2 Item #10 verifies the pressure in the SIT is between a high-high and a low pressure alarm setpoint. ITS SR 3.5.1.3 only verifies the pressure is above a minimum limit; whereas, STS SR 3.5.1.3 also defines an upper pressure limit.

Comment: DOC LA.3 states that neither the CTS (3.3.lb) nor the ITS specify a high pressure limit for the SITs since a maximum pressure is not an assumption in the safety analysis. However, the staff believes that, as evidenced by CTS Table 4.2.2 Item #10, the high pressure limit is an operability requirement for the SITs under the CTS. As the STS Bases state, the maximum nitrogen cover pressure limit ensures that excessive gas will not be injected into the RCS after the SITs have emptied. Please revise the.

submittal to retain the high pressure limit in conformance with the STS and the CTS.

Consumers Energy Response:

As stated in JFD 7 for Specification 3.5.1, "A maximum SIT nitrogen cover gas pressure is not assumed in any accident analysis for the Palisades plant. A limit on maximum gas pressure is administratively imposed to prevent accumulator relief valve actuation and ultimately preserve accumulator integrity." Testing for SIT level and pressure is specified by CTS 4.2, Table 4.2.2 item 10 which requires a verification "that level and pressure indication is between independent high high/low alarms for level and pressure." Although not clearly stated, the intent of this requirement is to ensure SIT volume is within the maximum and minimum amount (high level, low level alarms) assumed in the safety analysis, and that SIT pressure is greater than the minimum pressure (low pressure alarm) assumed in the safety analysis.

Since high SIT pressure is not an assumption in any accident analysis or a limitation imposed by CTS, it has not been incorporated in ITS 3.5.1.

Consistent with the response to NRC Request 3.5.G-1, DOC LA.3 has been deleted and replaced by DOC L.1 which justifies deletion of this CTS testing requirement.

Affected Submittal Pages:

See NRC Request 3.5.G-1 .

7

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.1-4 CTS 4.6.4.a ITS SR 3.5.1.5 JFD #6 CTS 4.6.4.a verifies each SIT flow path is Operable by verifying each motor-operated isolation valve is open and locking open the circuit breakers.

ITS SR 3.5.1.5 verifies power is removed from each SIT isolation valve operator.

Comment: STS SR 3.5.1.5 only requires this verification whenever the pressurizer pressure is greater than 2000 psia 11

Per JFD #6, this phrase was not retained in ITS SR 3.5.1.5 because of the change to the Applicability.

When the ITS 3.5.1 LCO Applicability is resolved (See Comment #3.5.1-1), this ITS SR may also need to be revised for consistency.

Consumers Energv Response:

See response to NRC Request 3.5.1-1 .

Affected Submittal Pages:

No page changes.

8

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS

NRC REQUEST:

3.5.2, ECCS - Operating 3.5.2-1 CTS 3.3.2.b, C, d, and f ITS 3.5.2 Action A DOC L.2 CTS 3.3.2.b, c, d, and fall require the separately li~ted LPSI, HPSI, shutdown cooling h~at exchanger and all valves associated with the safety injection and shutdown cooling system to be restored Operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ITS 3.5.2 Action A permits 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for any component that renders an ECCS train inoperable providing that 100% equivalent flow is still available from a, single train.

Comment #1: DOC L.2 as written only applies to CTS 3.3.2.c; whereas; the CTS markup indicates it is also applicable for CTS 3.3.2.b, d, and f. Revise this DOC to apply to all CTS changes.

Consumers Energv Response:

DOC L.1 has been revised to include CTS 3.3.2b, CTS 3.3.2d, and CTS 3.3.2f.

Affected Submittal Pages:

Att 3, DOC 3.5.2, pg 7 of 7 Att 4, NSHC 3.5.2, pg 2 of 4 Att 4, NSHC 3.5.2, pg 4 of 4

9

ATTACHNIENT 3 DISCUSSION OF CH~""JGES SPECIFICATION 3.5.2, ECCS - OPERA TING LESS RESTRICTIVE CHANGES (L)

L. l CTS 3.3.2 requires the reactor to be placed in cold shutdown within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months when an ECCS component is not restored to Operable status within the allowed outage time.

Proposed ITS 3. 5. 2 Required Action B. 2 requires the PCS temperature to be reduced to

< 325 °F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when an ECCS train is not restored to Operable status within the allowed outage time. Required Action B.2 places the plant in a condition in which two trains of ECCS are no longer necessary to mitigate the consequences of an accident assuming a single failure. In MODE 3 with PCS temperature < 325°F, ECCS operational requirements are relaxed due to stable plant conditions and the reduced probability of a OBA. ECCS requirements in MODE 3 with PCS temperature 1

< 325°F are addressed by proposed ITS 3.5.3, ECCS - Shutdown. Requiring the RM plant to be placed in a mode in which the LCO no longer applies is consistent with the ~.5'* ~-I philosophy of NUREG-1432. to/ti,. I C:tS .3.3 .z...b c. rs 3 .'J.2.c, I C7'.S JJ.2d. a.,.J C7S 3.3.U ft~uirc:. thc.ir /tsftc.+i'IJ( I f\oftrablt. CcCJ ~f L. 2 c; CTS 3. 3.;ie reqHires tl=lat aR iReperael8 WP~I pYITlp be restored to Operable status 'f...

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Proposed ITS Condition A has revised the CTS to allow one or more ECCS train to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least 100 3 of the ECCS flow equivalent to a single Operable ECCS train is available. Thus, the ITS allows aa IIPSI X subsy5tem, or a combination of HPSI and LPSI subsystem~to be inoperable for up to X j.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the flow assumed to be delivered by a single ECCS train is available. By stipulating a 1003 ECCS flow equivalent, proposed Condition A preserves the safety function of the ECCS system while allowing some period of time for correcting ECCS component inoperabilities. This allowance is acceptable because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. This change is supported by reliability analyses discussed in an NRC Memorandum to V. Stello, Jr., from R.L.

Baer, "Recommended Interim Revisions to LCOs for ECCS Components,"

December l, 1975. This change is consistent with NUREG-1432.

L3 f<AI 3.5. 5-/

Palisades Nuclear Plant Page 7 of 7 01120/98 9-()._/

ATTACH1\1ENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the plant is placed in a condition in which two ECCS trains are not required to function during an accident assuming a single failure of one train. Thus, this cQ.ange does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Prompt and appropriate actions have been determined based on safety analysis assumptions to ensure the plant is placed in a conditions under which two ECCS trains are no longer required assuming the single failure of one train.

Therefore, this change does not involve a significant reduction in the margin of safety.

~Al -'*~*i*I LESS RESTRICTIVE CHANGE L.2 C.t.WA .r J CTS J;,2. ~ e.n -'*32.c.) l'i$ ~~ zd °'"" C15 3.3.z.f' l?'(uirt -t~ir 1G5~+,ve.. lfltivero.blt cc.cs L.6mMd

( 4 cts 3.3.2e reE}\olires tkat aa i~1rablG BP~I pump be restored to Operable status within y 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Proposed. ITS Condition A has revised the CTS to allow one or more ECCS train to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least 1003 of the ECCS flow equivalent to a single Operable ECCS train is available. Thus, the ITS allows an HPSI il.lbiyitrm, er a combination k of HPSI and LPSI subsystems..Jto be* inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the flow assumed v to be delivered by a single ECCS train is available. By stipulating a 100% ECCS flow equivalent, proposed Condition A preserves the safety function of the ECCS system while allowing some period of time for correcting ECCS component inoperabilities. This allowance is acceptable because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. This change is supported by reliability analyses discussed in a NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components,"

December 1, 1975. This change is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 2 of 4 01/20/98 9-b

A TT ACHI\-lENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERA TING

~Al 3;z.. I

3. Does this change involve a significant reduction in a margin of safety? °"'"" 1 a.n 1no~ra.bk CCC5 Com !'<rrt The proposed change extends the allowed outage time for * (

1 from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and the combination of an HPSI ,, and LPSI from {

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provide at least 1003 flow equivalent to a single ECCS train is available. The function of the ECCS is to provide core cooling and negative reactivity to ensure the core is protected following various accidents. By stipulating a 1003 ECCS flow equivalent as a condition to extend the inoperability of an ECCS train, the safety function of the ECCS system is preserved. This is because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. As such, the margin of safety associated with the extended allowed outage time remains unchanged. Therefore, this change does not involve a significant reduction in a margin of safety.

AA/ 3.S.1-l.

L.3 f\k.uj C..om m "'° )..;.

See. ItJSc:RT

L.L( /J(.~

RAI 3 .5 .b-1 Sc..c.. :DJ'iltl1 x

Palisades Nuclear Plant Page 4 of 4 01120/98

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.2-1 CTS 3.3.2.b, c, d, and f ITS 3.5.2 Action A DOC L.2 CTS 3.3.2.b, c, d, and fall require the separately listed LPSI, HPSI, shutdown cooling heat exchanger and all valves associated with the safety injection and shutdown cooling system to be restored Operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ITS 3.5.2 Action A permits 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for any component that renders an ECCS train inoperable providing that 100% equivalent flow is still available from a single train.

Comment #2: The CTS 3.3.2.b and c markup states a second part to Condition B is added when this is not found in the ITS. Is this is an error and should the reference be to Condition A instead?

Consumers Energv Response:

The markup of CTS 3.3.2b and CTS 3.3.2c which states "Add Cond B Second Part" is in error and has been corrected to state "Add Cond A Second Part."

Affected Submittal Pages:

Att 3, CTS pg 3-29a (ITS 3.5.2, 2 of 5) 10

.3 EMERGENCY CORE CQQLING SYSTQ9 (Cont'd) 3.3.2

~Al 35.2.* I

b. Ont low-pressure s1f1ty 1njtct1on pumo ~1 ~ 1n CvM1 11 Z.

th* pu*p 1s restored to optr&blt st&tus within Obw~A c. Ont high-pressure Sifety injection pump *~Y bt i tht pump Is restored to operable status with In r-i.2wt"!::":":"::~-~~~-.-o1.

d.

3*29*

/()-a_,,

Amendment No. ~. 9-t, 172 September 26, 1996

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.2-2 CTS 3.3.1.e ITS 3.5.2 Action A DOC LA.1 CTS 3.3.1.e requires Operability of the shutdown cooling heat exchanger and all valves associated with the shutdown tooling system. ITS 3.5.2 simply requires that two ECCS trains be operable.

Comment: DOC LA.1 also applies to CTS 3.3.1.c, d, f, and g, and 3.3.4, which is acceptable. However, CTS 3.3.1.e requirements are not discussed in the Bases as stated in DOC LA.1. The inoperability of the shutdown cooling heat exchanger directly affects the ECCS train capability on recirculation flow. Bases insert to page B 3.5-10 and the LCO discussion do not provided complete description of the components in the ECCS trains that must be maintained Operable. Revise the ITS Bases accordingly to ensure these components are covered in ITS 3.5.2 and there is agreement with the CTS requirements relocated per DOC LA.1.

Consumers Energy Response:

The LCO portion of the Bases has been revised to discuss the operability requirements ass6ciated with the shutdown cooling heat exchange~s. The -

function of the shutdown cooling heat exchangers as they relate to the ECCS is to form part of the flow path that ensures adequate NPSH is available to the HPSI pumps during containment sump recirculation. Including these details in the Bases of ITS 3.5.2 ensures there is agreement between the requirement of CTS 3.3.le and the justification provided in DOC LA.1.

Affected Submittal Pages:

Att 2, ITS 3.5.2, pg B 3.5.2-6 Att 5, NUREG 3.5.2, pg B 3.5-13 Att 5, NUREG 3.5.2, pg B 3.5-14

-~---- ------ ------------~-~-

11

ECCS - Operating B 3.5.2

BASES LCO During an event requ1r1ng ECCS actuation, a flow path is (continued) provided to ensure an abundant supply of water from the SIRWT to the PCS, via the HPSI and LPSI pumps and their respective supply headers to each of the four cold le inJecti on nozz I e~ In t e o term, t is ow pa m e switc e o a e i s supply om the containment sump nd to

.!r.JSlRT supply rt of its flow to e PCS via loop 1 hot le drain

. e.

l in --**-----~----*-*--

/{Al 3 .S., '2.-1.

The flow path for each train must maintain its designed independence to ensure that no single active failure can disable both ECCS trains.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with PCS temperature

~ 325°F, the ECCS OPERABILITY requirements for the limiting Design Basis Accident (OBA) large break LOCA are based on full power operation .. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The HPSI pump performance is based on the small break LOCA, which establishes the pump performance curve and has less dependence on power. The requirements of MODE 2 and MODE 3 with PCS temperature ~ 325°F, are bounded by the MODE 1 analysis.

The ECCS functional requirements of MODE 3, with PCS temperature < 325°F, and MODE 4 are described in LCO 3.5.3, "ECCS - Shutdown."

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "PCS Loops - MODE 5, Loops Filled,"

and LCO 3.4.8, "PCS Loops - MODE 5, Loops Not Filled."

MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Shutdown Cooling (SOC) and Coolant Circulation - High Water Level," and LCO 3.9.5, "Shutdown Cooling (SOC) and Coolant Circulation - Low Water Level."

Palisades Nuclear Plant B 3.5.2-6 01/20/98 11-cv

INSERT During the recirculation phase, a flow path is provided from the containment sump to the PCS via the HPSI pumps. For worst case conditions, the containment building water level alone is not sufficient to assure adequate Net Positive Suction Head (NPSH) for the HPSI pumps.

Therefore, to obtain adequate NPSH, a portion of the Containment Spray (CS) pump discharge flow is diverted from downstream of the shutdown cooling heat exchangers to the suction of the HPSI pumps following recirculation during a large break LOCA. In this configuration, the CS pumps and shutdown cooling heat exchangers provide a support function for HPSI flow path OPERABILITY. The OPERABILITY requirements for the CS pumps and shutdown cooling heat exchangers are addressed in LCO 3.6.6, "Containment Cooling Systems."

Support system OPERABILITY is addressed by LCO 3.0.6 .

I 1-b

ECCS-Operat i ng B 3.5.2 BASES f'

APPLICABLE On smaller breaks,~S pressure will stabilize at a value SAFETY ANALYSES dependent upon breif size, heat load, and injection flow.

(continued) The smaller the breai, the higher this equilibrium pressure.

In all LOCA analyses, injection flow is not credited until PQtes pressure drops below the shutoff head of the HPSI pumps.

de.u~----'

~~~~~~~~~~~~-v.A~~~-,-~-oo-6~~~--.Pd,....,...-tc.-~-f-~~_,...+.Jr-,~~~3-ZS~'~F-,~~~

LCO In MODES IJ 2, and(3J'with ress izer r sure > 7 two independent (and redun ant) trains are required to ensure that sufficient ECCS flow is available, assuming there is a single failure affecting either train.

Additionally, individual coinponents within the ECCS trains may bt called upon to mitigate the consequences of other transients and accidents.

RA/ _3,5.2.-2.

(continued)

CEOG STS B 3.5-13 Rev 1, 04/07/95

!/-<:_

INSERT During the recirculation phase, a flow path is provided from the containment sump to the PCS via the HPSI pumps. For worst case conditions, the containment building water level alone is not sufficient to assure adequate Net Positive Suction Head (NPSH) for the HPSI pumps.

Therefore, to obtain adequate NPSH, a portion of the Containment Spray (CS) pump discharge flow is diverted from downstream of the shutdown cooling heat exchangers to the suction of the HPSI pumps following recirculation during a large break LOCA. In this configuration, the CS pumps and shutdown cooling heat exchangers provide a support function for HPSI flow path OPERABILITY. The OPERABILITY requirements for the CS pumps and shutdown cooling heat exchangers are addressed in LCO 3.6.6, "Containment Cooling Systems."

Support system OPERABILITY is addressed by LCO 3.0.6 .

I 1-J

ECCS-Operat 1ng B 3~5.2 BASES RA I .3. s. i. z.

The flow path for each train must ma1ntain its designed independence to ensure that no single~failure can disable both ECCS trains. ~~v~

p

~ APPLICABILITY In MODES 1 and 2-, and in MODE 3 with ~S(Qfes&um 1em ~ rwfvr't. =? '?>.2.S or:" C'.i:i7tio is!), the ECCS OPERABILITY requirements for the l1miting Oesign Basis Accident (OBA) large break LOCA are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The HPSI pump performance is based on the small break LOCA, which establishes th m nee curve and has less de endence on ow he charg er ance re u1r en s e ase The requirements o MOO 21 and 3, w1 th ~S ~re~....

r"i 2170o/i)s1il, are bounded by thelMODE 1 analys s.

~Ob~ .P The ECCS functional requirements of HOOE 3, with <<cs

~rei$ure <IJZOQ is!), and HOOE 4 are described in LCO 3.5.3,

ECCS-Shutdown.*

1 In HODES 5 and 6,~ conditions are such that the probability of an event requiring ECCS injection is extre11ely low. Core cooling requirements in HOOE Sare addressed by LCO 3.4.7, *Res Loops-MOOE 5, Loops Filled,*

and LCO 3.4.8, *Res Loops-HOOE 5, Loops Not Filled.*

HOOE 6 core cooling requirements are addressed by LCO 3.9.4,

shutdown Cooling (SOC) and Coolant Circulation-High Water Level,* and LCO 3.9.S, *shutdown Cooling (SOC) and Coolant Circulation-Lo~ Water Level.*

ACTIONS (continued)

CEOG STS B 3. 5-14 Rev 1, 04/07/95

J/-~

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS

NRC REQUEST:

3.5.2-3 CTS 3.3.2 ITS 3.5.2 Required Action B.2 DOC L.1 and JFD #15 CTS 3.3.2 allows an ECCS component to be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and then requires the reactor placed in a Cold Shutdown condition within an additional 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months (84 total) if the component cannot be restored. ITS Action A permits 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for restoring an inoperable ECCS train and an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (96 total) for placing the reactor in Mode 3 with PCS~ 325°F.

Comment: As discussed in JFD #15, a change to the Completion Time of Required Action B.2 may be appropriate to reach a lower operating temperature in Mode 3. However, no technical justification has been provided for the 24-hour Completion Time proposed for ITS 3.5.2, Required Action B.2. How does this Completion Time relate to similar Actions in other specifications requiring the plant to be placed in a similar operating condition? Please provide a justification for the 24-hour Completion Time for Required Action B.2.

Consumers Energv Response:

The Applicability of ITS 3.5.2 is Modes 1, 2, and Mode 3 with PCS temperature

~ 325°F. If the requirements of the LCO can not be met, the plant must be removed from the Modes or other specified conditions stated in the Applicability within a specified time. As such, the default action specified in Required Action B.2 is to reduce PCS temperature to < 325°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on operating experience to reach the required plant condition from full power in an orderly manner and without challenging plant system (JFD #15).

As discussed in DOC L.1, the default time specified in the ITS is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and the default time specified in the CTS is 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . Although the ITS default time is more restrictive than the CTS default time, the overall change has been characterized as less restrictive since the CTS places the plant in cold shutdown while the ITS only requires PCS temperature be reduced below 325°F .

12 (continued)

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.2-3 (continued)

Consumers Energv Response:

In general, default actions in the ITS which require the plant to be placed in Mode 4 specify a Completion Time of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Since the maximum average primary coolant temperature in Mode 4 is approximately 299°F, a plant cooldown from 532°F to 299°F in 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months would yield a cooldown rate of approximately 7.7°/hr. By comparison, a plant cooldown from 532°F to 325°F in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months would yield a cooldown rate of approximately 8.6°/hr. If a cooldown rate of 7.7°/hr were applied to Required Action B.2, the Completion Time would be approximately 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months . Consumers Energy considers a Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , although slightly more conservative, to be appropriate since it is consistent with other Completions Times specified in the ITS.

Affected Submittal Pages:

No page changes .

13

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.3, tees - Shutdown 3.5.3-1 No CTS ITS 3.5.3 DOC M.1; JFD #6 &#7 The CTS does not contain any ECCS requirements when the reactor is not critical. ITS 3.5.3 requires one LPSI train to be Operable in Mode 3 when PCS is < 325°F and in Mode 4; whereas, STS 3.5.3 requires one HPSI train Operable during Mode 3 when pressurizer pressure is < 1700 psia and in Mode 4.

Comment: This change in not consistent with either the CTS or the STS and is beyond the scope of the conversion review, and, therefore, will require additional review by the NRC technical staff. STS 3.5.3 requires at least one HPSI train operable during shutdown operations. To the contrary, ITS 3.5.3 requires that one LPSI train is required operable. JFD #6 states that this reflects the Palisades design. Is this difference solely due to LTOP? If so, why isn*t this explained in the Applicable Safety Analysis? The Applicable Safety Analysis Bases developed for ITS 3.5.3 directly state that the Bases for ITS 3.5.2 are applicable which still references analyses that credit only the HPSI pumps for the various sizes of LOCA breaks. There is no discussion or explanation regarding this difference from the STS. There is no discussion or explanation in the ITS 3.5.3 Bases for Applicable Safety Analyses pertaining to why one sole LPSI train is acceptable for the Palisades design.

ITS 3.5.2 Bases for Applicable Safety Analyses, paragraph 5, states that for all LOCA analyses, injection flow is not credited until PCS pressure drops below the shutoff head of the HPSI pumps. How do these analyses apply or how are they bounding for ITS 3.5.3 that now relies on only the LPSI pumps?

Revise the DOC and JFD, and ITS Bases to further explain and provide technical justification for changes to reflect the Palisades design.

Consumers Energy Response:

The Applicable Safety Analyses in the Bases for ITS 3.5.3 has been revised to better reflect the Palisade plant design. Specifically; discussions have been included to; 1) explain why the HPSI pumps are rendered incapable of injection when PCS temperature is < 325°F; 2) explain why one sole LPSI train is acceptable, 3) clarify that the acceptability for reduced ECCS operational requirements is based on engineering judgement rather than specific analysis, and 4) describe the analytical assumptions used to support one LPSI train in the shutdown conditions. In addition, the inappropriate statement that the 11 Applicable Safety Analyses section of Bases 3.5.2 is applicable to these Bases has also been deleted.

11 (continued) 14

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.3-1 (continued)

Consumers Energv Response:

In the Applicable Safety Analyses for Bases ITS 3.5.3 the statement which reads in all LOCA analyses, injection flow is not credited until PCS pressure 11 drops below the shutoff head of the HPSI pumps remains valid since the 11 shutoff head of a LPSI pump is below that of a HPSI pump. The analysis of record for a LOCA assumes the event is initiated at full power and is considered bounding for all credible events.

In response to the comment that there is no discussion or explanation 11 regarding the difference from the ITS that information is provided in JFD 6 11 for ITS 3.5.3.

Affected Submittal Pages:

Att 2, ITS 3.5.3, pg B 3.5.3-1 Att 5, NUREG 3.5.3, pg B 3.5-20

15

ECCS - Shutdown B 3.5.3

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS - Shutdown BASES BACKGROUND The Background section for Bases B 3.5.2, ECCS - Operating,"

11 is applicable to these Bases, with the following modifications. '

In MODE 3 with Primary Coolant System (PCS) temperature

< 325°F and in MODE 4, an ECCS train is defined as one Low Pressure Safety Injection (LPSI) train. The LPSI flow path consists of piping, valves, and pumps that enable water from the Safety Injection Refueling Water Tank (SIRWT), and subsequently the containment sum, to be in"ected in o the PCS following he ac i en s escr1 e in B a Lbl5 o~ Golo.r.i Ac.c.1d""'"'+ C.Lo'-'4}. X APPLICABLE Bases 3.5.2 /{1+1 SAFETY ANALYSES .)_s.3- I

:CNSutT ~

ue to the stable ca ditions associa ed with operatic ODE 3 with PCS te erature < 325°F and in MODE 4, a the educed probabili of a Design Ba is Accident (DBA), the CCS operational equirements are educed. Includ in hese reduction is that certain automatic safety njection

\

ignals are no available. In ese conditions, ufficient RAI 3.S.3-1 ime exists fa manual actuati n of the require ECCS omponents t mitigate the co sequences of a D .

nly one tr. in of ECCS is r quired in MODE 3 ith PCS

. emperatun < 325°F and in MODE 4. Protect"on against ingle f lures is not re ied on for this DE of operation.

ECCS - Shutdown satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO In MODE 3 with PCS temperature < 325°F and in MODE 4, an ECCS train is comprised of a single LPSI train. Each LPSI train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the SIRWT and transferring suction to the containment sump.

Palisades Nuclear Plant B 3.5.3-1 01/20/98 I~-()_,

ECCS-Shutdown s* 3. 5 .3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 8 3. 5. 3 ECCS-Shutdown RC.S ~ rts R.uJ r . . S!R(..Jj BASES HP .i't : lis-t BACKGROUND The Background section for Bases B 3. 5. 2, "ECCS-Operat i ng,

is applicable to these Bases, with the following modi fie at ions.

.4\Ma.ry ~'t ~ '"""" A " ~ 3lS' F-

+t'"'c.,.... ffljection:-<IBPSil,iS"ubiyrlenn Thet.>SI f1ow path consists of piping, valves, and pum s that enable water from the

.:5c..fc.1'1 ~ ~/:fueling .Jrat!!:_,hnk be injected into the i:w~..,ea~e.....m,~

<t"l!Q1ant/$yStEm~CSt( owin e accidents described in Bases 3. 5. 2. P (.s1(w1')

,o.nd ~~j\)(..n11f -tht. (t,r\to.11'1 mVI~ 01.J"" f, --------

Af\ \

~.s.~-l

~

EC~~ttdown satisfies Criterion 3 of@; HR!t Polity)

(Sh . / D c.Ft. So.~ Cc.. t2.)

.;_ fCS ~""tt.,""'1f<. < ~Z.S O? and 1n ffiCl:k'f LCO In HOOE 3 with fpreriyri zet pressllf.e < peo pS/j I>, an EC~S 'if'a,"

(~~~is composed of a single~SI subsystem. Eactr4!PSI

-tf'c..1,., includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the~ and transferring suction to the containment sump.

rs,~

(continued)

CEOG STS B 3.5-ZO Rev 1, 04/07/95 I 5-b

B 3 5 3 INSERT 1 In Mode 3 with PCS temperature< 325°F and in Mode 4 the normal compliment ofECCS components is reduced from that which is available during operations above Mode 3 with PCS temperature ~ 325 °F. The acceptability for the reduced ECCS operational requirements is based on engineering judgement rather than specific analysis and considers such factors as the reduced probability that a LOCA will occur, and the reduced energy stored in the fuel. The reduction in ECCS operational requirements include:

1) Isolation of the Safety Injection Tanks (SITs) since PCS pressure is expected to be reduced below the SIT injection pressure,

2) Reliance on manual safety injection initiation since the automatic Safety Injection Signal (SIS) is not required by the technical specifications below 300°F,

3) Rending the High Pressure Safety Injection (HPSI) pumps incapable of injecting into the PCS. The HPSI pumps are rendered incapable of injecting into the PCS in accordance with the requirements of LCO 3. 4 .12, "Low Temperature Overpressure Protection (LTOP) System". This action assures that a single mass addition event initiated at a pressure within the limits ofLCO 3.4.12 cannot cause the PCS pressure to exceed the 10 CPR 50 Appendix G limit.

At a PCS temperature of325°F the maximum allowed PCS pressure corresponds to the LTOP setpoint limit which is approximately 800 psia. Below 800 psia postulated piping flaws of critical size are considered unlikely since normal operation at 2060 psia serves as a proof test against ruptures. In addition, since the reactor has been shutdown for a period oftime, the decay heat and sensible heat levels are greatly reduced from the full power case.

Although a pipe break in the PCS pressure boundary is considered unlikely, break sizes larger and smaller than approximately 0. 1fl:2 are considered separately when analyzing ECCS response.

For breaks larger than approximately O. lfl:2, the event is characterized by a very rapid depressurization of the PCS to near the containment pressure. Due to the reduced temperature and pressure of the PCS, the time to complete blowdown is extended from that assumed in the full power case. During this time, the fuel is cooled by the flow through the core towards the break. Automatic safety injection actuation is not assumed to occur since the pressurizer pressure SIS maybe bypassed below 1700 psig. Therefore, operator action is relied upon to initiate ECCS flow. Indication that would alert the operator that a LOCA had occurred include; a loss of pressurizer level, rapid decrease in PCS pressure, increase in containment pressure, and containment high radiation alarm. Since the saturation pressure for 325 °F is approximately 100 psia, the LPSI pumps are capable of providing the required heat removal function. When the OPERABLE LPSI pump is being used to fulfill the shutdown cooling function, the PCS pressure is< 300 psia. As such, the rate of PCS blowdown is reduced providing some time to manually realign the OPERABLE LPSI pump to the ECCS mode of operation.

For breaks smaller than approximately O. l:ft:2, the event is characterized by a slow depressurization of the PCS and a relatively long time for the PCS level to drop below the tops of the hot legs. In MODE 3 with PCS temperature< 325°F and in the upper range of MODE 4 before shutdown cooling is established, the spectrum of smaller break sizes are more limiting than larger breaks in terms ofECCS performance since the PCS could stay above the shutoff head of the LPSI pumps. For these break sizes, sufficient time, well in excess ofrecommended 10 minutes attributed for manual operator action, is available to either initiate once through cooling using the PORVs, or by re-establishing HPSI pump injection capability. In either case, the core remains covered and the criteria of 10 CFR 50.46 preserved .

/6-d

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.3-2 CTS 3.3.1 ITS SR 3.5.3.1 JFD #9 CTS 3.3.1 provides no specific requirements for determining the Operability of the ECCS system component during the shutdown modes of Operation. ITS SR 3.5.3.1 provides direct reference to those surveillances from ITS 3.5.2 that are applicable to the Operability requirements of ITS 3.5.3.

Conunent: Based upon the information contained in the Bases for ITS 3.5.2 and ITS 3.5.3, the STS SRs 3.5.2.1, 3.5.2.3, 3.5.2.6, 3.5.2.7~ and 3.5.2.9 are still applicable to ITS 3.5.3 as follows: (1) The ESF Pump Mini Flow valves (CV-3027 and CV-3056) preclude damage to all ECCS pumps (as stated in Bases inserts 1, 2, and 3 on page B 3.5-16) and not just to the HPSI. Therefore, the position of these valves should be verified per SR 3.5.2.1. (2) The ECCS pumps and valves are automatically actuated per the design which does cease to exist, even though the slower evolution of the design basis event in these Modes happens a rate that permits a manual response. The design has not changed; therefore, STS SRs 3.5.2.6 and 3.5.2.7 should apply, as indicated in the STS. (3) Similar to the previous item, the LPSI flowpaths should be verified as Operable with the stops in the correct position. In addition for both items (2) and (3), in order to enter normal power operation mode for ITS 3.5.2, SR 3.0.4 requires that these SRs must be met prior to entry into the Modes of Applicability; so, it will be necessary to fulfill these SR requirements during the Applicability of ITS 3.5.3 to ensure the safety function is met prior to entering the Applicability of ITS 3.5.2. The staff requests that you revise the ITS to adopt these STS SR requirements.

Consumers Ener<<v Response: .

CV-3027 and CV-3056 are in the Safety Injection Refueling Water Tank (SIRWT) recirculation flow path which provides protection for the ECCS pumps when the pumps are aligned for ECCS injection. In Mode 4, the LPSI pumps are also used for shutdown cooling which requires the pump suctions be realigned from the SIRWT to the PCS. To preclude having a direct flow path from the PCS to the SIRWT, CV-3027 and CV-3056 must be maintained closed. Since the ECCS components required by ITS 3.5.3 rely on manual actuation and thus, do not require the protection afforded by the SIRWT recirculation path, and to ~void a direct flow path from the PCS to the SIRWT, the requirement of ISTS SR 3.5.2.1 have not been incorporated in ITS 3.5.3.

(continued) 16

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS NRC REQUEST:

3.5.3-2 (continued)

Consumers Energy Response:

Major assumptions of ITS 3.5.3 is that the required ECCS components are actuated manually and that no credit is taken for any safety injection signal.

The basis for these assumptions is that the safety injection function (automatic and manual initiation) is not required to be Operable in Mode 4.

As such, over 90% of the Applicable temperature range of ITS 3.5.3 (200°F to 325°F) is in a condition where automatic safety injection is not required.

Thus, it is not necessary to require each ECCS automatic valve actuate to its correct position on an actual or simulated actuation signal(ISTS SR 3.5.2.6),

or that each ECCS pump starts automatically on an actual or simulated action signal (ISTS SR 3.5.3.7) when the initiation signal is not required to be Operable. Assurance the required valves can be positioned as necessary, and that the ECCS pumps can be started, is provided by ASME Section XI testing required by 10 CFR 50.55a.

A verification that each ECCS throttle valve "position stop" is in th~ correct position ensures the ECCS is capable of delivering the flow rate assumed in the safety analysis. This is achieved by ensuring the valve is properly positioned such that flow through the valve body is optimized. The LOCA analysis assumes the event is initiated near full power conditions and that ECCS flow rates are influenced by the elevated containment pressure which acts to restrict flow. For events initiated at PCS pressures within the Applicability of ITS 3.5.3, the increases in containment pressure will be significantly less than the full power case. As such, the influence on ECCS flow rates will be reduced, providing sufficient flow rate margin. Therefore, the requirement of ISTS SR 3.5.2.9 has not been incorporated in ITS 3.5.3. To better clarify the basis of the ECCS throttle valve position stops, a revision has been made to the Bases of SR 3.5.2.8 to describe the operation and design of the HPSI valve limit stops and LPSI valve limit stops.

While it is recognized SR 3.0.4 would preclude entry into LCO 3.5.2 until all Applicable SRs have been met, it is also recognized these SRs have a Frequency of 18 months. From an operational perspective, these SRs would typically be performed during a refueling outage and would remain current throughout plant operations in Mode 4. However, the decision not to include these SRs in ITS 3.5.3 is based on the fact these SRs are not necessary to assure the requirements of LCO 3.5.3 are met.

Affected Submittal Pages:

Att 2, ITS 3.5.2, pg B 3.5.12-12 Att 5, NUREG 3.5.2, pg B 3.5-18 17

ECCS - Operating B 3.5.2

BASES SURVEILLANCE REQUIREMENTS SR 3.5.2.8 (continued) Realignment of alves in the flow path on n SIS is

~Al necessary for roper ECCS performance. Te open limit 3,5,3-a.. switch on ea of the four LPSI cold leg isolation valves and the two PSI hot leg isolation valv s is set to

.INS:RT ~ establish predetermined flow which e sures that a singl low press e safety injection subsyst m is capable of

veri the flow rate re uired in the safet anal si .

The 18 month Frequency 1s ase on e same ac ors as stated above for SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7.

SR 3.5.2.9 Periodic inspection of the containment sump ensures that it is unrestricted and stays in proper operating condition.

The 18 month Frequency is based on the need to perform this Surveillance under outage conditions. This Frequency is sufficient to detect abnormal degradation and is confirmed by operating experience.

REFERENCES 1. FSAR, Section 5.1

2. FSAR, Section 14.17

3. NRC Memorandum to V. Stello, Jr., from R. L. Baer, 11 Recorranended Interim Revisions to LCOs for ECCS Components," December l, 1975

4. IE Information Notice No. 87-01, January 6, 1987

Palisades Nuclear Plant B 3.5.2-12 04/09/99

ECCS-Operat i ng B 3.5.2 BASES SURVEILLANCE SB 3.5.2.5 REQUIREMENTS (continued)

.....-v SB 3.5 .. 2j' SB 3~5.2J@and SB 3.s.2$P These SBs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated

~---.s165 @I'§ jb A RAS, that each ECCS pump starts on receipt of

~ an actual or simulated S~, and that the LPSI pumps stop on rece1p o an ac ua or simulated RAS. This Surveil~ance is o u re or va ves a are locked, sealed, or otherwise secured in the required posit ion under administrative controls. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also accepta ed on consideration of the desi n reliability L )'(and operating ex eri enc e e u1 . The on o c s es e as par o t e Engineered Safety Feature

st (ES~) testing, and equipment performance is mon to e as part of the Inservice Testing Progrui.

~~!s. J"7- SB 3.5.2.©V

~.,.._~~-t-----:-~~~~~~~~~----__,

(-I~-~-~-Tf---t (continued)

CEOG STS B 3.5-18 Rev 1, 04/07/95

INSERT SR 3.5.2.8 The HPSI Hot Leg Injection motor operated valves and the LPSI loop injection valves have position switches which are set at other than the full open position. This surveillance verifies that these position switches are set properly.

The HPSI Hot leg injection valves are manually opened during the post-LOCA long term cooling phase to admit HPSI injection flow to the PCS hot leg. The open position limit switch on each HPSI hot leg isolation valves is set to establish a predetermined flow split between the HPSI injection entering the PCS hot leg and cold legs.

The LPSI loop injection MOVs open automatically on a SIS signal. The open position limit.

switch on each LPSI loop injection valve is set to establish the maximum possible flow through that valve. The design of these valves is such that excessive turbulence is developed in the valve body when the valve disk is at the full open position. Stopping the valve travel at slightly less than full open reduces the turbulence and results in increased flow. Verifying that the position stops are properly set ensures that a single low pressure safety injection subsystem is capable of delivering the flow rate required in the safety analysis .

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION

NRC REQUEST:

3.5.4 SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS Safety Injection Refueling Water Tank (SIRWT)

No comments.

3.5.5 Trisodium Phosphate (TSP)

No comments .

18

ENCLOSURE 2 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION EDITORIAL CHANGES

SITs B 3.5.1

BASES APPLICABLE The 1720 ppm limit for minimum boron concentration was SAFETY ANALYSES established to ensure that, following a LOCA with a minimum (continued) level in the SITs, the reactor will remain subcritical in the cold condition following mixing of the SITs, Safety Injection Refueling Water Tank and PCS water volumes. Small break LOCAs assume that al~control rods are inserted, except or the control roa of highest worth, which is withdrawn from the core. Large break LOCA analyses assume

,that all~control rods remain withdrawn until the blowdown phase is over. For large break LOCAs, the initial reactor shutdown is accomplished by void formation. The most limiting case occurs at beginning of core life.

The maximum boron limit of 2500 ppm in the SITs is based on boron precipitation in the core following a LOCA. With the reactor vessel at saturated conditions, the core dissipates heat by boiling. Because of this boiling phenomenon in the core, the boric acid concentration will increase in this region. If allowed to proceed in this manner, a point will be reached where boron precipitation will occur in the core.

Post LOCA emergency procedures direct the operator to establish simultaneous hot and cold leg injection to prevent this condition by establishing a forced flow path through the core regardless of break location. These procedures are based on the minimum time in which precipitation could occur, assuming that maximum boron concentrations exist in the borated water sources used for injection following a LOCA. Boron concentrations in the SITs in excess of the limit could result in precipitation earlier than assumed in the analysis.

The SITs satisfy Criterion 3 of 10 CFR 50.36(c)(2).

LCO The LCO establishes the minimum conditions required to ensure that the SITs are available to accomplish their core cooling safety function following a LOCA. Four SITs are required to be OPERABLE to ensure that 100% of the contents of three of the SITs will reach the core during a LOCA.

This is consistent with the assumption that the contents of one tank spill through the break. If the contents of fewer than three tanks are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 could be

violated.

Palisades Nuclear Plant B 3.5.1-5 01/20/98

ECCS - Operating B 3.5.2 BASES BACKGROUND Each train of a Safety Injection Signal (SIS) actuates LPSI (continued) flow by starting one LPSI pump and opening two LPSI loop injection valves. Each train of an SIS actuates HPSI flow by starting one HPSI pump, opening the four associated HPSI loop injection valves, and closing the pressure control Confi rrno.torj valves associated with each Safety Injection Tank. In add1t1on, each train of a SIS will provide ~pen signal to the~omponent Cooling Water valves which supply seal and

---....b-ea-Jri ng cooling to the LPSI, HPSI, and Containment Spray pumps.

The safety analyses assume that one only train of safety injection is available to mitigate an accident. While operating under the provisions of an ACTION, an additional single failure need not be assumed in assuring that a loss of function has not occurred. Therefore, the LPSI flow assumed in the safety analyses can be met if there is an OPERABLE LPSI flow path from the SIRWT to any two PCS loops.

The HPSI flow assumed in the safety analyses can be met if there is an OPERABLE HPSI flow path from the SIRWT to each PC~ ~ee~. In each case, an OPER~flow path must include an OPERABLE pump and an OPERABLE~ injection valve.

A suction header supplies water from the SIRWT or the containment sump to the ECCS pumps. Separate piping supplies each train. The discharge headers from each HPSI pump divide into four supply lines after entering the containment, one feeding each PCS cold leg. The discharge headers from each LPSI pump combine to supply a common header which divides into four supply lines after entering containment, one feeding each PCS cold leg.

Motor operated valves are set to maximize the LPSI flow to the PCS. This flow balance directs sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the PCS cold legs.

For LOCAs coincident with a loss of off-site power that are too small to initially depressurize the PCS below the shutoff head of the HPSI pumps, the core cooling function is provided by the Steam Generators (SGs) until the PCS pressure decreases below the HPSI pump shutoff head .

Palisades Nuclear Plant B 3.5.2-2 01/20/98

SIRWT B 3.5.4

BASES APPLICABLE Twenty minutes is the point at which approximately 75% of SAFETY ANALYSES the design flow of one HPSI pump is capable of meeting or (continued) exceeding the decay heat boiloff rate.

The SIRWT capacity, alone, is not sufficient to provide adequate Net Positive Suction Head (NPSH) for the HPSI pu~p~

after switch over to the containment sump for the worst case conditions. To assure adequate NPSH for the HPSI pumps, their suction headers are aligned to the discharge of the Containment Spray Pumps (Ref. 2). Restrictions are placed on Containment Spray Pump operation with this alignment to ensure the Containment Spray Pumps have adequate NPSH (Ref. 3).

In MODE 4, the minimum volume limit of 200,000 gallons is based on engineering judgement and considers factors such as:

a. The volume of water transferred from the SIRWT to the PCS to account for the change in PCS water volume during a cooldown from 532°F to 200°F {approximately 17,000 gallons assuming an initial PCS volume of 80,000 gallons); and

b. The minimum SIRWT water volume capable of providing a sufficient level in the containment sump to support LPSI pump operation following a LOCA.

Due to the reduced PCS temperature and pressure requirements in MODE 4, and in recognition that water from the SIRWT used for PCS makeup is available for recirculation following a LOCA, the minimum water volume limit for the SIRWT in MODE 4 is lower than in MODES 1, 2, or 3.

The 1720 ppm limit for minimum boron concentration was established to ensure that, following a LOCA with a minimum level in the SIRWT, the reactor will remain subcritical in the cold condition following mixing of the SIRWT, Safety Injection Tanks, and PCS water volumes. Small break LOCAs .:i.-'

assume that all control rods are inserted, except for the r:S!.

contra ro a highest worth, which is withdrawn from the core. Large break LOCA analyses assume that alJ,control ro s remain aw un 1 neoTowaown phase is over. For large break LOCAs, the initial reactor shutdown is accomplished by void formation. The most limiting case

occurs at beginning of core life.

Palisades Nuclear Plant B 3.5.4-3 01/20/98

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.4, REFUELING WATER TANK M.3 CTS 4.17, Table 4.17 .5 item #12 specifies a Channel Check of the SIRW Tank level instrumentation. Proposed ITS SR 3.5.4.2 requires a verification that the SIRWT borated water volume is <!: 250,000 gallons. The CTS definition of a Channel Check states, "A Channel Check shall include verification that the monitored parameter is within limits imposed by the Technical Specifications." As such, the Channel Check requirement of CTS 4.17 encompasses the level verification requirement of ITS SR 3. 5 .4. 2. The frequency associated with the Ohannel Check of CT S 4 .17 is every 92 days. The frequency associated with the level verification of ITS SR 3.5.4.2 is every 7 days. As such, the requirement of proposed ITS SR 3.5.4.2 is more restrictive than the requirement of CTS 4.17. Althoagh a ChaBnel Cheek ef SIR'.VT leoel c:;d jostrument wm stm be required every 92 da~'ii iR the ITS ~~88 SeetiBH 3.3), tire -

defimtioR gf Cb.ainiel Cb.eek 120 longer requires a verificatioR tb.at tR.e F&enitere'd parameter is within the Technical Sped~atiQR liH:lit. This change is consistent with NUREG-1432.

LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

LA.1 CTS 4.17, Table 4.17 .6 item 3 specifies the requirement to perform a Channel Check and Channel Calibration on the SIRW Tank temperature instrumentation. Although SIRW Tank temperature is assumed in the plant safety analysis, the instrumentation associated with this parameter is not credited in the analysis since it does not provide a mitigative or protective function. The assumption in the safety analysis is that the SIRW Tank temperature is within the assumed temperature band. As such, the method for determining SIRW Tank temperatures can be located outside of the technical specifications without a significant impact on safety. Therefore, the CTS requirement to perform a Channel Check and Channel Calibration on the SIRW Tank temperature instrumentation is being moved to the Operating Requirements Manual. Placing this information in the Operating Requirements Manual provides adequate assurance that the surveillance requirements will be maintained. Changes to the Operating Requirements Manual are subject to the provisions of 10 CFR 50.59. This change is consistent with the removal of .similar type instruments from the STS (e.g.,

accumulator level and pressure instruments addressed in NUREG-1366, "Improvements to Technical Specification Requirements") and NUREG-1432.

Palisades Nuclear Plant Page 3 of 4 01120/98

ENCLOSURE 3 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS Page Change Instructions Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.

REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT#

ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL No page changes.

ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL ITS B 3.5.1-5 ITS B 3.5.1-5 04/09/99 editorial ITS B 3.5.2-2 ITS B 3.5.2-2 04/09/99 editorial ITS B 3.5.2-6 ITS B 3.5.2-6 04/09/99 RAI 3.5.2-2 ITS B 3.5.2-7 ITS B 3.5.2-7 04/09/99 -----------

ITS B 3.5.2-8 ITS B 3.5.2-8 04/09/99 -----------

ITS B 3.5.2-9 ITS B 3.5.2-9 04/09/99 -----------

ITS B 3.5.2-10 ITS B 3.5.2-10 04/09/99 -----------

ITS B 3.5.2-11 ITS B 3.5.2-11 04/09/99 -----------

ITS B 3.5.2-12 ITS B 3.5.2-12 04/09/99 -----------

ITS B 3.5.2-13 ITS B 3.5.2-13 04/09/99 RAI 3.5.3-2 ITS B 3.5.3-1 ITS B 3.5.3-1 04/09/99 RAI 3.5.3-1 ITS B 3.5.3-2 ITS B 3.5.3-2 04/09/99 RAI 3.5.3-1 ITS B 3.5.3-3 ITS B 3.5.3-3 04/09/99 RAI 3.5.3-1

ITS B 3.5.3-4 04/09/99 -----------

ITS B 3.5.3-5 04/09/99 -----------

ITS B 3.5.4-3 ITS B 3.5.4-3 04/09/99 editorial ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL CTS 3.5.1, pg 3-29a CTS 3.5.1, pg 3-29a 04/09/99 RAI 3.5.1-2 RAI 3.5.2-1 CTS 3.5.1, pg 4-11 CTS*3.5.l, pg 4-11 04/09/99 RAI 3.5.G-1 CTS 3.5.1, pg 4-24 CTS 3.5.1, pg 4-24 04/09/99 RAI 3.5.G-1 CTS 3.5.2, pg 3-29a CTS 3.5.2, pg 3-29a 04/09/99 RAI 3.5.1-2 RAI 3.5.2-1 CTS 3.5.2, pg 4-24 CTS 3.5.2, pg 4-24 04/09/99 RAI 3.5.G-1 DOC 3.5.1, pg 2 of 5 DOC 3.5.1, pg 2 of 5 04/09/99 RAI 3.5.1-2 DOC 3.5.1, pg 3 of 5 DOC 3.5.1, pg 3 of 5 04/09/99 RAI 3.5.G-1 DOC 3.5.1, pg 4 of 5 DOC 3.5.1, pg 4 of 5 04/09/99 -----------

DOC 3.5.1, pg 5 of 5 DOC 3.5.1, pg 5 of 5 04/09/99 RAI 3.5.G-1 RAI 3.5.1-2 DOC 3.5.2, pg 6 of 7 DOC 3.5.2, pg 6 of 7 04/09/99 RAI 3.5.G-1 RAI 3.5.2-1

DOC 3.5.2, pg 7 of 7 DOC 3.5.4, pg 3 of 4 DOC 3.5.2, pg 7 of 7 DOC 3.5.4, pg 3 of 4 1

04/09/99 04/09/99 RAI 3.5.G-l RAI 3.5.1-2 editorial

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO MARCH 17, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS REMOVE PAGES 1

INSERT PAGES REV DATE NRC COMMENT#

ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL NSHC 3.5.1, pg 1 of 3 NSHC 3.5.1, pg 1 of 7 04/09/99 NSHC 3.5.l, pg 2 of 3 NSHC 3.5.1, pg 2 of 7 04/09/99 NSHC 3.5.1, pg 3 of 3 NSHC 3.5.1, pg 3 of 7 04/09/99

NSHC 3.5.l, pg 4 of 7 04/09/99 RAI 3.5.1-2

NSHC 3.5.1, pg 5 of 7 04/09/99 RAI 3.5.1-2

NSHC 3.5.1, pg 6 of 7 04/09/99 RAI 3.5.G-1

NSHC 3.5.1, pg 7 of 7 04/09/99 RAI 3.5.G-1 NSHC 3.5.2, pg 1 of 4 NSHC 3.5.2, pg 1 of 7 04/09/99 NSHC 3.5.2, pg 2 of 4 NSHC 3.5.2, pg 2 of 7 04/09/99 RAI 3.5.2-1 NSHC 3.5.2, pg 3 of 4 NSHC 3.5.2, pg 3 of 7 04/09/99 NSHC 3.5.2, pg 4 of 4 NSHC 3.5.2, pg 4 of 7 04/09/99 RAI 3.5.1-2

NSHC 3.5.2, pg 5 of 7 04/09/99 *RAI 3.5.1-2

NSHC 3.5.2, pg 6 of 7 04/09/99 RAI 3.5.G-1 RAI 3.5.1-2 NSHC 3.5.2, pg 7 of 7 04/09/99 RAI 3.5.G-1 ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG B 3.5-13 NUREG B 3.5-13 04/09/99 RAI 3.5.2-2

NUREG B 3.5-13 insert 04/09/99 RAI 3.5.2-2 NUREG B 3.5-14 NUREG B 3.5-14 04/09/99 RAI 3.5.2-2 NUREG B 3.5-18 NUREG B 3.5-18 04/09/99 RAI 3.5.3-2 NUREG B 3.5-18 insert NUREG B 3.5-18 insert 04/09/99 RAI 3.5.3-2 NUREG B 3.5-20 NUREG B 3.5-20 04/09/99 RAI 3.5.3-1

NUREG B 3.5-20 insert (2 pgs) 04/09/99 RAI 3.5.3-1 ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL No page changes.

2

SI Ts B 3.5.1 BASES APPLICABLE The 1720 ppm limit for minimum boron concentration was SAFETY ANALYSES established to ensure that, following a LOCA with a minimum (continued) level in the SITs, the reactor will remain subcritical in the cold condition following mixing of the SITs, Safety Injection Refueling Water Tank and PCS water volumes. Small break LOCAs assume that all full-length control rods are inserted, except for the control rod of highest worth, which is withdrawn from the core. Large break LOCA analyses assume that all full-length control rods remain withdrawn until the blowdown phase is over. For large break LOCAs, the initial reactor shutdown is accomplished by void formation. The most limiting case occurs at beginning of

core life.

The maximum boron limit of 2500 ppm in the SITs is based on boron precipitation in the core following a LOCA. With the reactor vessel at saturated conditions, the core dissipates heat by boiling. Because of this boiling phenomenon in the core, the boric acid concentration will increase in this region. If allowed to proceed in this manner, a point will be reached where boron precipitation will occur in the core.

Post LOCA emergency procedures direct the operator to establish simultaneous hot and cold leg injection to prevent this condition by establishing a forced flow path through the core regardless of break location. These procedures are based on the minimum time in which precipitation could occur, assuming that maximum boron concentrations exist in the borated water sources used for injection following a LOCA. Boron concentrations in. the SITs in excess of the limit could result in precipitation earlier than assumed in the analysis.

The SITs satisfy Criterion 3 of 10 CFR 50.36(c)(2).

LCO The LCO establishes the minimum conditions required to ensure that the SITs are available to accomplish their core cooling safety function following a LOCA. Four SITs are required to be OPERABLE to ensure that 100% of the contents of three of the SITs will reach the core during a LOCA.

This is consistent with the assumption .that the contents of one tank spill through the break. If the contents of fewer than three tanks are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 could be

violated.

Palisades Nuclear Plant B 3.5.1-5 04/09/99

ECCS - Operating B 3.5.2

BASES BACKGROUND (continued)

Each train of a Safety Injection Signal (SIS) actuates LPSI flow by starting one lPSI pump and opening two LPSI loop injection valves. Each train of an SIS actuates HPSI flow by starting one HPSI pump, opening the four associated HPSI loop injection valves, and closing the pressure control valves associated with each Safety Injection Tank. In addition, each train of a SIS will provide a confirmatory open signal to the normally open Component Cooling Water valves which supply seal and bearing cooling to the LPSI, HPSI, and Containment Spray pumps.

The safety analyses assume that one only train of safety

injection is available to mitigate an accident. While operating under the provisions of an ACTION, an additional single failure need not be assumed in assuring that a loss of function has not occurred. Therefore, the LPSI flow assumed in the safety analyses can be met if there is an OPERABLE LPSI flow path from the SIRWT to any two PCS loops.

The HPSI flow assumed in the safety analyses can be met if there is an OPERABLE HPSI flow path from the SIRWT to each cold leg. In each case, an OPERABLE flow path must include an OPERABLE pump and an OPERABLE injection valve .

A suction header supplies water from the SIRWT or the containment sump to the ECCS pumps. Separate piping supplies each train. The discharge headers from each HPSI pump divide into four supply lines after entering the containment, one feeding each PCS cold leg. The discharge headers from each LPSI pump combine to supply a common header which divides into four supply lines after entering containment, one feeding each PCS cold leg.

Motor operated valves are set to maximize the LPSI flow to the PCS. This flow balance directs sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the PCS cold legs.

For LOCAs coincident with a loss of off-site power that are too small to initially depressurize the PCS below the shutoff head of the HPSI pumps, the core cooling function is provided by the Steam Generators (SGs) until the PCS pressure decreases below the HPSI pump shutoff head.

Palisades Nuclear Plant B 3.5.2-2 04/09/99

ECCS - Operating B 3.5.2

BASES LCO (continued)

During an event requ1r1ng ECCS actuation, a flow path is provided to ensure an abundant supply of water from the SIRWT to the PCS, via the HPSI and LPSI pumps and their respective supply headers, to each of the four cold leg injection nozzles is available. During the recirculation phase, a flow path is provided from the containment sump to the PCS via the HPSI pumps. For worst case conditions, the containment building water level alone is not sufficient to assure adequate Net Positive Suction Head (NPSH) for the HPSI. pumps. Therefore, to obtain adequate NPSH, a portion of the Containment Spray (CS) pump discharge flow is diverted from downstream of the shutdown cooling heat exchangers to the suction of the HPSI pumps following recirculation during a large break LOCA. In this configuration, the CS pumps and shutdown cooling heat exchangers provide a support function for HPSI flow path OPERABILITY. The OPERABILITY requirements for the CS pumps and shutdown cooling heat exchangers are addressed in LCO 3.6.6, "Containment Cooling Systems." Support system OPERABILITY is addressed by LCO 3.0.6 *

The flow path for each train must maintain its designed independence to ensure that no single active failure can disable both ECCS trains.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with PCS temperature 2 325°F, the ECCS OPERABILITY requirements for the limiting Design Basis Accident (OBA) large break LOCA are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The HPSI pump performance is based on the small break LOCA, which establishes the pump performance curve and has less dependence on power. The requirements of MODE 2 and MODE 3 with PCS temperature 2 325°F, are bounded by the MODE 1 analysis.

The ECCS functional requirements of MODE 3, with PCS temperature< 325°F, and MODE 4 are described in LCO 3.5.3, "ECCS - Shutdown."

Palisades Nuclear Plant B 3.5.2-6 04/09/99

ECCS - Operating B 3.5.2 BASES APPLICABILITY In MODES 5 and 6, plant conditions are such that the (continued) probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "PCS Loops - MODE 5, Loops Filled,"

and LCO 3.4.8, "PCS Loops - MODE 5, Loops Not Filled."

MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Shutdown Cooling (SOC) and Coolant Circulation - High W,ater Level," and LCO 3.9.5, "Shutdown Cooling (SOC) and Coolant Circulation - Low Water Level."

ACTIONS If one or more trains are inoperable, but at least 100% of the flow assumed to be delivered by a single OPERABLE ECCS train is available, the inoperable components must be returned to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on an NRC study (Ref. 3) using a reliability evaluation and is a reasonable amount of time to effect many repairs.

An ECCS train is inoperable if it is not capable of delivering the assumed flow to the PCS. The individual components are inoperable if they are not capable of performing their required function, or if supporting systems are not available.

The safety analyses assume that only one train of safety injection is available to mitigate an accident. While operating under the provisions of an ACTION, an additional single failure need not be assumed in assuring that a loss of function has not occurred. Therefore, the LPSI flow assumed in the safety analyses can be met if there is an OPERABLE LPSI flow path from the SIRWT to any two PCS loops.

The HPSI flow assumed in the safety analyses can be met if there is an OPERABLE HPSI flow path from the SIRWT to each PCS loop. In each case, an OPERABLE flow path must include an OPERABLE pump and an OPERABLE loop injection valve.

Palisades Nuclear Plant B 3.5.2-7 04/09/99

ECCS - Operating B 3.5.2 BASES ACTIONS A.1 (continued)

The LCO requires the OPERABILITY of two independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of this Condition is to maintain a combination of OPERABLE equipment such that 100% of the ECCS flow assumed to be delivered by a single OPERABLE train remains available. This allows increased flexibility in plant operations when components in opposite trains are inoperable.

An event accompanied by a loss of offsite power and the failure of an emergency DG can disable one ECCS train until power is restored. A reliability analysis (Ref. 4) has shown that the impact with one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Reference 4 describes situations in which one component, such as the shutdown cooling flow control valve, CV-3006, can disable both ECCS trains. With one or more components inoperable, such that 100% of the flow assumed to be delivered by a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be immediately entered.

B.1 and B.2 If the inoperable train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and PCS temperature reduce to < 325°F 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 in an orderly manner and without challenging plant systems. *

Palisades Nuclear Plant B 3.5.2-8 04/09/99

ECCS - Operating B 3.5.2

BASES SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the PCS is maintained. CV-3027 and CV-3056 are stop valves in the minimum recirculation flow path for the ECCS pumps. If either of these valves were closed when the PCS pressure was above the shutoff head of the ECCS pumps, the pumps could be damaged by running with insufficient flow and thus render both ECCS trains inoperable.

Placing HS-3027A and HS-30278 for CV-3027, and HS-3056A and HS-30568 for CV-3056, in the open position* ensures that the valves cannot be inadvertently misaligned or change position as the result of an active failure. These valves are of the type described in Reference 4, which can disable the function of both ECCS trains and invalidate the accident analysis. CV-3027 and CV-3056 are capable of being closed from the control room since the SIRWT must be isolated from the containment during the recirculation phase of a LOCA. A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered reasonable in view of other administrative controls ensuring that a mispositioned valve is an unlikely possibility.

SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS 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 that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position .

Palisades Nuclear Plant 8 3.5.2-9 04/09/99

ECCS - Operating B 3.5.2

BASES SURVEILLANCE SR 3.5.2.2 (continued)

REQUIREMENTS The 31 day Frequency is appropriate because the valves are operated under procedural control and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience.

SR 3.5.2.3 SR 3.5.2.3 verifies CV-3006 is in the open position and that its air supply is isolated. CV-3006 is the shutdown cooling flow control valve located in the common LPSI flow path.

The valve must be verified in the full open position to support the low pressure injection flow assumptions used in the accident analyses. The inadvertent misposition of this valve could result in a loss of low pressure injection flow and thus invalidate these flow assumptions. CV-3006 is designed to be held open by spring force and closed by air pressure. To ensure the valve cannot be inadvertently misaligned or change position as the result of a hot short in the control circuit, the air supply to CV-3006 is isolated. Isolation of the air supply to CV-3006 is acceptable since the valve does not require repositioning during an accident.

The 31 day Frequency has been shown to be acceptable through operating practice and the unlikely occurrence of the air supply to CV-3006 being unisolated coincident with a inadvertent valve misalignment event or a hot short in the control circuit.

Palisades Nuclear Plant B 3.5.2-10 04/09/99

ECCS - Operating B 3.5.2

BASES SURVEILLANCE SR 3.5.2.4 REQUIREMENTS (continued) Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller damage or other hydraulic component problems is required by Section XI of the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the Inservice Testing Program, which encompassesSection XI of the ASME Code.Section XI of the ASME Code provides the activities and Frequencies necessary to satisfy the requirements.

SR 3.5.2.5. SR 3.5.2.6. and SR 3.5.2.7 These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated actuation signal, i.e., on an SIS or RAS, that each ECCS pump starts on receipt of an actual or simulated actuation signal, i.e., on an SIS, and that the LPSI pumps stop on receipt of an actual or simulated actuation signal, i.e., on an RAS. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month Frequency i~ based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power .. The 18 month Frequency is also acceptable based on consideration of the design reliability of.the equipment and operating experience. The actuation logic is tested as part of the Engineered Safety Feature (ESF) testing, and equipment performance is monitored as part of the Inservice Testing Program .

Palisades Nuclear Plant B 3.5.2-11 04/09/99

ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.8 REQUIREMENTS (continued) The HPSI Hot Leg Injection motor operated valves and the LPSI loop injection valves have position switches which are set at other than the full open position. This surveillance verifies that these position switches are set properly.

The HPSI Hot leg injection valves are manually opened during the post-LOCA long term cooling phase to admit HPSI injection flow to the PCS hot leg. The open position limit switch on each HPSI hot leg isolation valves is set to establish a predetermined flow split between the HPSI injection entering the PCS hot leg and cold legs.

The LPSI loop injection MOVs open automatically on a SIS signal. The open position limit switch on each LPSI loop injection valve is set to establish the maximum possible flow through that valve. The design of these valves is such that excessive turbulence is developed in the valve body when the valve disk is at the full open position. Stopping the valve travel at slightly less than full open reduces the turbulence and results in increased flow. Verifying that the position stops are properly set ensures that a single low pressure safety injection subsystem is capable of delivering the flow rate required in the safety analysis.

The 18 month Frequency is based on the same factors as those stated above for SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7.

SR 3.5.2.9 Periodic inspection of the containment sump ensures that it is unrestricted and stays in proper operating condition.

The 18 month Frequency is based on the need to perform this Surveillance under outage conditions. This Frequency is sufficient to detect abnormal degradation and is confirmed by operating experience. *

Palisades Nuclear Plant B 3.5.2-12 04/09/99

ECCS - Operating B 3.5.2

BASES REFERENCES 1. FSAR, Section 5.1

2. FSAR, Section 14.17

3. NRC Memorandum to V. Stello, Jr., from R. L. Baer, 11 Recommended Interim Revisions to LCOs for ECCS Components, 11 December l, 1975

4. IE Information Notice No. 87-01, January 6, 1987

Palisades Nuclear Plant B 3.5.2-13 04/09/99

ECCS - Shutdown B 3.5.3

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS - Shutdown BASES BACKGROUND The Background section for Bases B 3.5.2, "ECCS - Operating,"

is applicable to these Bases, with the following modifications.

In MODE 3 with Primary Coolant System (PCS) temperature

< 325°F and in MODE 4, an ECCS train is defined as one Low Pressure Safety Injection (LPSI) train. The LPSI flow path consists of piping, valves, and pumps that enable water from the Safety Injection Refueling Water Tank (SIRWT), and subsequently the containment sump, to be injected into the PCS following a Loss of Coolant Accident (LOCA).

APPLICABLE In Mode 3*with PCS temperature< 325°F and in Mode 4 the SAFETY ANALYSES normal compliment of ECCS components is reduced from that which is available during operations above Mode 3 with PCS

temperature ~ 325°F. The acceptability for the reduced ECCS operational requirements is based on engineering judgement rather than specific analysis and considers such factors as the reduced probability that a LOCA will occur, and the reduced energy stored in the fuel. The reduction in ECCS operational requirements include:

1) Isolation of the Safety Injection Tanks (SITs) since PCS pressure is expecte~ to be reduced below the SIT injection pressure,

2) Reliance on manual safety injection initiation since the automatic Safety Injection Signal (SIS) is not required by the technical specifications below 300°F, Palisades Nuclear Plant B 3.5.3-1 04/09/99

ECCS - Shutdown B 3.5.3 BASES APPLICABLE 3) Rending the High Pressure Safety Injection (HPSI)

SAFETY ANALYSES pumps incapable of injecting into the PCS. The HPSI (continued) pumps are rendered incapable of injecting into the PCS in accordance with the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

System". This action assures that a single mass addition event initiated at a pressure within the limits of LCO 3.4.12 cannot cause the PCS pressure to exceed the 10 CFR 50 Appendix. G limit.

At a PCS temperature of 325°F the maximum allowed PCS pressure corresponds to the LTOP setpoint limit which is approximately 800 psia. Below 800 psia postulated piping flaws of critical size are considered unlikely since normal operation at 2060 psia serves as a proof test against ruptures. In addition, since the reactor has been shutdown for a period of time, the decay heat and sensible heat levels are greatly reduced from the full power case.

Although a pipe break in the PCS pressure boundary is considered unlikely, break sizes larger and smaller than approximately 0.1 ft 2 are considered separately when analyzing ECCS response.

For breaks larger than approximately 0.1 ft 2 , the event is characterized by a very rapid depressurization of the PCS to near the containment pressure. Due to the reduced temperature and pressure of the PCS, the time to complete blowdown is extended from that assumed in the full power case. During this time, the fuel is cooled by the flow through the core towards the break. Automatic safety injection actuation is not assumed to occur since the pressurizer pressure SIS maybe bypassed below 1700 psig.

Therefore, operator action is relied upon to initiate ECCS flow. Indication that would alert the operator that a LOCA had occurred include; a loss of pressurizer level, rapid decrease in PCS pressure, increase in containment pressure, and containment high radiation alarm. Since the saturation pressure for 325°F is approximately 100 psia, the LPSI pumps are capable of providing the required heat removal function.

When the OPERABLE LPSI pump is being used to fulfill the shutdown cooling function, the PCS pressure is < 300 psia.

As such, the rate of PCS blowdown is reduced providing some time to manually realign the OPERABLE LPSI pump to the ECCS mode of operation.

Palisades Nuclear Plant B 3.5.3-2 04/09/99

ECCS - Shutdown B 3.5.3

BASES APPLICABLE For breaks smaller than approximately 0.1 ft 2 , the event is SAFETY ANALYSES characterized by a slow depressurization of the PCS and a (continued) relatively long time for the PCS level to drop below the tops of the hot legs. In MODE 3 with PCS temperature

< 325°F and in the upper range of MODE 4 before shutdown cooling is established, the spectrum of smaller break sizes are more limiting than larger breaks in terms of ECCS performance since the PCS could reach saturation above the shutoff head of the LPSI pumps. For these break sizes, sufficient time, well in excess -of recommended 10 minutes attributed for manual operator action, is available to either initiate once through cooling using the PORVs, or by re-establishing HPSI pump injection capability. In either case, the core remains covered and the criteria of 10 CFR 50.46 preserved.

ECCS - Shutdown satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCD In MODE 3 with PCS temperature < 325°F and in MODE 4, an ECCS train is comprised of a single LPSI train. Each LPSI train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from I

the SIRWT and transferring suction to the containment sump.

During an event requiring ECCS actuation, a flow path is required to supply water from the SIRWT to the PCS via one LPSI pump and at least one supply header to a cold leg injection nozzle. In the long term, this flow path may be switched to take its supply from the containment sump.

With PCS temperature < 325°F, one LPSI pump is acceptable without single failure consideration, based on the stable reactivity condition of the reactor and the limited core cooling requirements. The High Pressure Safety Injection (HPSI) pumps may therefore be released from the ECCS train requirements. With PCS temperature< 300°F, both HPSI pumps must be rendered incapable of injection into the PCS in accordance with LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

Palisades Nuclear Plant B 3.5.3-3 04/09/99

ECCS - Shutdown B 3.5.3

BASES LCO The LCO is further modified by a Note that allows a LPSI (continued) train to be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned (remote or local) to the ECCS mode of operation and not otherwise inoperable. This allows operation of a LPSI pump in the shutdown cooling mode.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with PCS temperature

~ 325°F, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 3 with PCS temperature < 325°F and in MODE 4, one OPERABLE ECCS train is acceptable without single failure consideration, based on the stable reactiv~ty condition of the reactor and the limited core cooling requirements.

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "PCS Loops - MODE 5, Loops Filled,"

and LCO 3.4.8, "PCS Loops - MODE 5, Loops Not Filled."

.MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Shutdown Cooling (SOC) and Coolant Circulation - High Water Level," and LCO 3.9.5, "Shutdown Cooling (SOC) and Coolant Circulation - Low Water Level."

ACTIONS With no LPSI train OPERABLE, the plant is not prepared to respond to a loss of coolant accident. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time to restore at least one LPSI train to OPERABLE status ensures that prompt action is taken to restore the required cooling capacity.

When the Required Action cannot be completed within the required Completion Time, a controlled shutdown should be initiated. Twenty-four hours is reasonable, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems.

Palisades Nuclear Plant B 3.5.3-4 04/09/99

ECCS - Shutdown B 3.5.3 BASES SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply.

REFERENCES The applicable references from Bases 3.5.2 apply.

Palisades Nuclear Plant B 3.5.3-5 04/09/99

SIRWT B 3.5.4

BASES APPLICABLE SAFETY ANALYSES Twenty minutes is the point at which approximately 75% of the design flow of one HPSI pump is capable of meeting or (continued) exceeding the decay heat boiloff rate.

The SIRWT capacity, alone, is not sufficient to provide adequate Net Positive Suction Head (NPSH) for the HPSI pumps after switch over to the containment sump for the worst case conditions. To assure adequate NPSH for the HPSI pumps, their suction headers are aligned to the discharge of the Containment Spray Pumps (Ref. 2). Restrictions are placed on Containment Spray Pump operation with this alignment to ensure the Containment Spray Pumps have adequate NPSH (Ref. 3).

In MODE 4, the minimum volume limit of 200,000 gallons is based on engineering judgement and considers factors such as:

a. The volume of water transferred from the SIRWT to the PCS to account for the change in PCS water volume during a cooldown from 532°F to 200°F (approximately 17,000 gallons assuming an initial PCS volume of 80,000 gallons); and

b. The minimum SIRWT water volume capable of providing a sufficient level in the containment sump to support LPSI pump operation following a LOCA.

Due to the reduced PCS temperature and pressure requirements in MODE 4, and in recognition that water from the SIRWT used for PCS makeup is available for recirculation following a LOCA, the minimum water volume limit for the SIRWT in MODE 4 is lower than in MODES 1, 2, or 3.

The 1720 ppm limit for minimum boron concentration was established to ensure that, following a LOCA with a minimum level in the SIRWT, the reactor will remain subcritical in the cold condition following mixing of the SIRWT, Safety Injection Tanks, and PCS water volumes. Small break LOCAs assume that all full-length control rods are inserted, except for the control rod of highest worth, which is withdrawn from the core. Large break LOCA analyses assume that all full-length control rods remain withdrawn until the blowdown phase is over. For large break LOCAs, the initial reactor shutdown is accomplished by void formation. The most limiting case occurs at beginning of core life.

Palisades Nuclear Plant B 3.5.4-3 04/09/99

3.3 3.3.2 EMERGENCY CORE CQQLING SYSTEM (Cont'd) ts o are ~

a11 be p1aced in a ~

One safety injection tank *ay bt inoperable for a period of no more]~~

ttan one hour. ~

b. One ow* ressure sa ety injectio pump may be inoperable r

s. ., )

the pu is restored to operabl status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

( 3,S." c. igh-pressure safety inje tion PWllP may be inoperab i restored to o er bl* status within 4 h r .

v a. -..;.;;.~~~~~~~~:-:::t~-:7~~~=--'

G~~\)~~ ny r *__e~'---......-..---..---.----..--...----ir.--~--~--=----:

er oc irect y associated with one of Ste.) the above co to function during ace dent

( 3-1 conditions put of that component and shall

.. ,t th* s isttd for that componen .

f. Any valve interlock or pip associated with the saf ty injection and shut own cooling syste a~d which is not covere under 3.3.2e above b , which is reQui d to function during ac ident conditi ns, ~ay bt inope ble for a period of no re than

@I

< AGD G,NlJ. A) cw

( f1DD C.otJJ) D) 8 3*29&

Amendment Ho. ~. i-t, 172 September 26. 1996 Rev;sed .

04/09/99 ;z, of 5

Amend1111nt No*. *' a.I-, m, ffi, Hi, ~, ~'

4-11

.. . Rev;sed 04/09/99

j .s.}

4. 6 @EEIY INJECJIOH,f/AAD 4.6.1 COHTAINt!E~ SPRAY SYST::4S ,fEs§J-

1. t ts shill be perfonned each reactor refueli g interval.

A test fety injection signal w 11 be applied to init' te operat

n of the syste111. The s fety injection and sh down cool ~rig syste pump motors may be de-e ergized for this test. The system will e considered satisfacto y if control board ind cation and vis 1 observations indicate that all components ha e received the ty injection signal in he proper sequence and timing (ie, tl1e a ropriatt puap breaKers all have opened and osed, and all alves shall have CO!lplet d their travel .

I I

1. SysteN test shall be perfol"'11ed at each reactor refuelin9 lnterva .

The test shall be perfonMJd w1th the 1solat1on valves In the s ay supply lines at the contain~nt blocKed closed. Operation of the syste11 is initiated by tr pin9 the nonnal actuation ins.truMntation.

At 1ust s tht spray nozzles open *

4.6.3 The test will indicate all considtrtd s1tisf1ctory if visual observations

ponents hive operated satisfacto ly.

1. The injection PU"'PS, shut own cooling pumps, and s shall be started at intervals not to exceed

(

s~c.

J*f*l. )

months Alternate Nanu1l star ing between control room the l cal bruKer shill bt pr cticed In the test progn.

and

3. '- b. table levels of perfo ance shall be. that the pups start, re h their rated heads o recirculation flow, and o erate for at 1st fifteen 1inut1s.

4.6.4 y11 yu 1.

S~3S.l.I.

r,~.S.I. 5 The Lo Pressure S1f1 y Injection flo path shall be erified OPERAS E within 7 d s prior to eac reactor startu by verifying flow ontrol valve open, and its air su ply is isolate<: l .

4*24 Amendment Ho. i+,

Rev;sed

~. 9-i, W, ffi, ~. 174 October 31, l 996 04/09/99 5 af S

3.5.~

EMERGENCY CORE CQQLING SYSTEM (Cont'd) 3.3.2 During power oper1tion t

requ remen s o 11 w one of the followin conditions to bt f C.orJb r>.i /6.z.i~:.5~!;~~d ~P;~; f~:~ ~;row ~:-:--;;~:-r.:..:..;~,..;.:~..:.;,~~:..;..,;=-=-=-~~.:.:......:t::..h:..:.e_~

S.fa.ncM.yfou@o!!))condition within not niet w1thin 1n 1ddit;on1

~

Sc'-

3 5.t

b. Ont low-pressure s1fety injection pump **Y be in

~N!JA tht pump is restored to oper1bl1 st1tus within GiwtlA c. One high-pressure s1fety Injection pump m1y be i tht pu~ is restored to oper1blt st1tus within~~~=-~~

d.

~

3.3.2e

3*291 Amendment No. ~. fH., 172 Rev;sed September 26, 1996 04/09/99

3.S.,G

4.6 SAFETY IHJECTIQN AHO CQNTAIMf1EHT SPRAY SYSTEMS TESTS 4.6.1

.1:-----@

SR.3.5.?..5 SR..3.5".2.~

4.6.2

1. Sys e~ test shall be perfon1e at each re1ctor refueling in rval.

T test*sh1ll be performed 1th the isolation valves int spray pply lines at the contai nt blocKed closed. Operatic of the

\

( ~.~)

s~e. ystet1 is initiated by tri ping the normal actuation instruMntltion.

At least sha 11 be v open *

4.6.3 The test will be c sidertd satisfactory if visua observations indicate all comp ents have operated satisfacto ily.

~

JR ~.5.2*'-/

a. (St.<.)

3~

SR *;J.s.c,.I.(

2 b* Acee t1ble levels of - ~erfonaance shall befth~lie pumps start, reac e r r1ti01ie1 s on recirculation flow, and operate for at

( '5'<<-o.l<o 3 *I..) 1east f 1ftetn...lin.\il.l.S_. ____..... -****-*-------------......:------**-

4. 6. 4 v1Jyu

&. S1f y Injection Tank flow path shall be verifi d OPERABLE r actor st1rtup by ver ying each

( See)

~.S./

oper1t 1so11tion valve is pen by observing val indic ion and v1lve itself and locKing open th associated circ t brtaKers.

position

b. be verified 5 R 3.S.Z3 o tac reac or s ar u y ver1 y1ng open, and its air supply is isolated .

f2*Q 4-24 Amendment Ho. i-t-, ~. %, W, ffi, ~. 174 October 31, 1996 Revised 04/09/99 o~ s

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

A.4 CTS 3.3.1 has been revised to include a new Condition, Required Action and Completion Time to address two or more inoperable SITs. The proposed change

. requires an immediate entry into LCO 3. 0. 3. This proposed change does not impose any new requirement on plant operation since the CTS currently requires entry into LCO 3.0.3 with two or more inoperable SITs. The intent of this change is to eliminate the potential for the misinterpretation of the Required Actions if two or more SITs were inoperable for different reasons (i.e., boron concentration not within limits in one SIT concurrent with the inability to verify pressure or level in another SIT). Since this change only impacts the presentation of the CTS requirements, it is considered administrative in nature.

A.5 CTS 3.3.2e provides the required actions for any valves, interlocks or piping directly associated with the SITs. CTS 3.3.2f provides the required actions for any valves, interlocks or piping associated with the SITs which is not covered by CTS 3.3.2e. In the ITS, these same valves, interlocks and pipes are addressed by the definition of Operability. An earlier version of the CTS defined Operable as "a system or component is operable if it is capable of fulfilling its design function.~" As such, to ensure all the components necessary to fulfill the safety function of a system, subsystem or train were adequately covered, the CTS provided two distinct required actions .

Subsequent changes to the CTS have redefined the term "Operable" such that it is no longer necessary to provide two different required actions. Since the revised definition of Operable in the CTS is consistent with the definition of Operable in the ITS, the required actions of CTS 3.3.2e and CTS 3.3.2f are no longer necessary.

MORE RESTRICTIVE CHANGES (M)

M.1 CTS 3 .3 .1 contains an exception which allows the reactor to be made critical for low temperature physics testing even though all the conditions associated with the SITs are not met. Proposed ITS 3 .5 .1 does not contain this same exception and, as such, always requires the SITs to be Operable prior to entering MODE 2. This change is an additional restriction on plant operations and is consistent with NUREG-1432 .

- Palisades Nuclear Plant Page 2 of 5 04/09/99

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

M.2 CTS 3.3.2 requires the reactor to be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when one SIT is not restored to Operable status within the allowed outage time. Proposed ITS 3.5.1 Required Action C.1 requires the plant to be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months when one SIT is not restored to Operable status within the allowed outage time.

This change has been proposed to establish consistency within the ITS for the time necessary to reach MODE 3. This change continues to ensure that a plant shutdown can be achieved in a controlled manner with the least impact on plant systems likely to induce an undesired plant transients. This change is an additional restriction on plant operations and is consistent with NUREG-1432.

M.3 In CTS 4.6.4a, the frequency specified for position verification of the SIT motor operated isolation valves and their associated breakers is "within 7 days prior to each reactor startup." Proposed ITS SR 3.5.1.1 requires a verification of each SIT isolation valve every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Proposed SR 3.5.1.5 requires a verification that power is removed from each SIT isolation valve every 31 days. Proposed SR 3.0.1 states that failure to meet a surveillance, whether such failure is experienced during the performance of the surveillance or between performance of the surveillance, shall be failure to meet the LCO. Proposed SR 3.0.4 states that entry into a Mode or other specified condition in the Applicability of an LCO shall not be made unless the LCO's surveillances have been met within their specified Frequency. As such, the ITS imposes additional restrictions relative to the frequency at which the SRs associated with the SIT motor operated isolation valves are performed. This change is consistent with NUREG-1432.

LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

There were no "Removal of Details" changes associated with this Specification .

Palisades Nuclear Plant Page 3 of 5 04/09/99

ATTACHMENT 3

LESS RESTRICTIVE CHANGES (L)

DISCUSSION OF CHANGES SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

L.1 CTS 3.3.2 requires the reactor to be placed in cold shutdown when one SIT is not restored to Operable status within the allowed outage time. Proposed ITS 3.5.1 Required Action C.1 requires the plant to be placed in MODE 3 when one SIT is not restored to Operable status within the allowed outage time. Required Action C. l places the plant in a condition in which the SITs are no longer necessary to mitigate the consequences of an accident. In MODE 3 and below, the rate of PCS blowdown during a LOCA is such that the HPSI and LPSI pumps can provide adequate injection to ensure peak cladding temperature remains below the 10 CFR 50 .46 limit of 2200 °F.

Only requiring the plant to be placed in a mode in which the LCO does not apply is consistent with the philosophy of NUREG-1432.

L.2 A new Condition, Required Action and Completion Time is proposed for the case where one SIT is inoperable due to its boron concentration not being within limits.

This addition in the ITS allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore boron versus the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the CTS.

The S~Ts are passive devices and the boron concentration is relatively stable.

Therefore, deviations from the normal boron concentration band are generally small and the impact on a LOCA is minor. This is because boiling of the emergency core cooling water during the reflood phase of a LOCA concentrates the boron in the saturated liquid that remains in the core. In addition, the volume of the SIT is still available for injection. Since the boron requirements are based on the average boron concentration of the total volunie of three SITs, the consequences are less severe than they would be if an SIT were not available for injection.

Palisades Nuclear Plant Page 4 of 5 04/09/99

- L. 3 SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

ATTACHMENT 3 DISCUSSION OF CHANGES CTS 3 .3. 2 contains a provision which allows one of the ECCS components required by CTS 3. 3 .1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements of LCO 3.0.3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss of ECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a "condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss of ECCS function will not occur. This is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432 .

L.4 CTS 4.6.4a and CTS 4.2, Table 4.2.2 item #10 contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion. Specifically, CTS 4.6.4a states that SIT motor operated isolation valve position is verified "by observing valve position indication and (the) valve itself." In addition, CTS 4.6.4a also requires a verification that the breakers for the SIT motor operated isolation valves are locked opened. The intent of verifying valve position and removing power from the valve operator is to ensure that the valve is in its correct position and that an active failure does not result in the undetected closure of a valve. Specifying that valve position be verified "by observing valve position indication and (the) valve itself', or stating that the "breaker must be locked opened" do not constitute requirements assumed in the safety analyses. Rather, they simply provide a method for assuring valve and breaker position. CTS 4.2 Table 4.2.2 item 10 contains a surveillance requirement to verify SIT pressure is less than the high alarm. The purpose of this alarm is to alert the operator of an off normal condition (e.g., nitrogen inleakage or check valve back leakage). This requirement has been deleted since the safety analyses do not assume a maximum SIT pressure. Thus, it is not appropriate to contain a requirement whose function is strictly to alert the operator of an off normal condition when the Operability of the affected component is not impacted. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 5 of 5 04/09/99

ATTAC1'ENT 3

LA.1 (continued)

DISCUSSION OF CHANGES SPECIFICATION 3.5.2, ECCS - OPERATING Removing the details of the ECCS from the CTS and placing them in the Bases of the ITS is acceptable since these details are not pertinent to the actual requirements.

Placing these details in the Bases provides adequate assurance that they will be maintained since the Bases are controlled by the Bases Control Program proposed in ITS Chapter 5.0. This change is consistent with NUREG-1432.

LESS RESTRICTIVE CHANGES (L)

L.1 CTS 3.3.2 requires the reactor to be placed in cold shutdown within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months when an ECCS component is not restored to Operable status within the allowed outage time.

Proposed ITS 3.5.2 Required Action B.2 requires the PCS temperature to be reduced to

< 325°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when an ECCS train is not restored to Operable status within the allowed outage time. Required Action B.2 places the plant in a condition in which two trains of ECCS are no longer necessary to mitigate the consequences of an accident assuming a single failure. In MODE 3 with PCS temperature < 325°F, ECCS operational requirements are relaxed due to stable plant conditions and the reduced probability of a DBA. ECCS requirements in MODE 3 with PCS temperature

< 325°F are addressed by proposed ITS 3.5.3, ECCS - Shutdown. Requiring the plant to be placed in a mode in which the LCO no longer applies is consistent with the philosophy of NUREG-1432.

L.2 CTS 3.3.2.6, CTS 3.3.2c, CTS 3.3.2d, and CTS 3.3.2f require their respective inoperable ECCS component be restored to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Proposed ITS Condition A has revised the CTS to allow one or more ECCS train to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least 100% of the ECCS flow equivalent to a single Operable ECCS train is available. Thus, the ITS allows a combination of HPSI and LPSI subsystems to be inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the flow assumed to be delivered by a single ECCS train is available. By stipulating a 100% ECCS flow equivalent, proposed Condition A preserves the safety function of the ECCS system while allowing some period of time for correcting ECCS component inoperabilities.

This allowance is acceptable because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. This change is supported by reliability analyses discussed in an NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975. This change is consistent with

NUREG-1432.

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ATTACHMENT 3

L.3 DISCUSSION OF CHANGES SPECIFICATION 3.5.2, ECCS - OPERATING CTS 3.3.2 contains a provision which allows one of the ECCS components required by CTS 3.3.1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements of LCO 3.0.3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss of ECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a"condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss of ECCS function will not occur. This is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432 .

L.4 CTS 4.6. la and CTS 4.6.3a contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion. Specifically, CTS 4.6.lA contains the requirement for testing the Safety Injection System. The CTS states that satisfactory results of the test are indicated by "control board indication and visual observation." The intent of this test is to verify that components which receive an acttiation signal actuate to their correct position.

Specifying that valve position be verified "by "control board indication and visual observation." does not constitute a requirement assumed in the safety analyses. Rather, it simply provide a method for assuring valve position. CTS 4.6.3a states that "alternate manual starting (of the safety injection pumps and shutdown cooling pumps) between the control room console and the local breaker shall be practiced in the test program." The ability to demonstrate the manual starting capability of the ECCS pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capability of meeting their intended safety function. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432.

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ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.5.4, REFUELING WATER TANK M.3 CTS 4.17, Table 4.17 .5 item #12 specifies a Channel Check of the SIRW Tanlc level instrumentation. Proposed ITS SR 3.5.4.2 requires a verification that the SIRWT borated water volume is ~ 250,000 gallons. The CTS definition of a Channel Check states, "A Channel Check shall include verification that the monitored parameter is within limits imposed by the Technical Specifications." As such, the Channel Check requirement of CTS 4.17 encompasses the level verification requirement of ITS SR 3. 5 .4. 2. The frequency associated with the Channel Check of CTS 4 .17 is every 92 days. The frequency associated with the level verification of ITS SR 3.5.4.2 is every 7 days. As such, the requirement of proposed ITS. SR 3. 5 .4. 2 is more restrictive than the requirement of CTS 4.17. This change is consistent with NUREG-1432.

LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUl\fENTS (LA)

LA.1 CTS 4.17, Table 4.17 .6 item 3 specifies the requirement to perform a Channel Check and Channel Calibration on the SIRW Tanlc temperature instrumentation. Although SIRW Tanlc temperature is assumed in the plant safety analysis, the instrumentation associated with this parameter is not credited in the analysis since it does not provide a mitigative or protective function. The assumption in the safety analysis is that the SIRW Tank temperature is within the assumed temperature band. As such, the method for determining SIRW Tank temperatures can be located outside of the technical specifications without a significant impact on safety. Therefore, the CTS requirement to perform a Channel Check and Channel Calibration on the SIRW Tank temperature instrumentation is being moved to the Operating Requirements Manual. Placing this information in the Operating Requirements Manual provides adequate assurance that the surveillance requirements will be maintained. Changes to the Operating Requirements Manual are subject to the provisions of 10 CFR 50.59. This change is consistent with the removal of similar type instruments from the STS (e.g., accumulator level and pressure instruments addressed in NUREG-1366, "Improvements to Technical Specification Requirements") and NUREG-1432 .

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ATTAC1'ENT 4

LESS RESTRICTIVE CHANGE L.1 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

CTS 3.3.2 requires the reactor to be placed in cold shutdown when one SIT is not restored to Operable status within the allowed outage time. Proposed ITS 3.5.1 Required Action C. l requires the plant to be placed in MODE 3 when one SIT is not restored to Operable status within the allowed outage time. Required Action C. l places the plant in a condition in which the SITs are no longer necessary to mitigate the consequences of an accident. In MODE 3 and below, the rate of PCS blowdown during a LOCA is such that the HPSI and LPSI pumps can provide adequate injection to ensure peak cladding temperature remains below the 10 CFR 50.46 limit of 2200°F. Only requiring the plant to be placed in a mode in which the LCO does not apply is consistent with the philosophy of NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change limits the required condition in which the plant is placed when one SIT can not be restored to Operable to below the condition where the SITs are credited in the safety analysis. By placing the plant in a condition which the SITs are no longer necessary to mitigate the consequences of an accident there can be no impact on the probability or consequences of an accident previously evaluated. Therefore the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the plant is placed in a condition in which the SITs are not assumed to function during an accident.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

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

The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change.

Prompt and appropriate actions have been determined based on safety analysis assumptions to ensure the plant is placed in a conditions under which the SIT safety function is no longer required. Therefore, this change does not involve a significant reduction in the margin of safety.

LESS RESTRICTIVE CHANGE L.2 A new Condition, Required Action and Completion Time is proposed for the case where one SIT is inoperable due to its boron concentration not being within limits, or one SIT inoperable due to the inability to verify level or pressure. This addition in the ITS allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore boron versus the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the CTS. The SITs are passive devices and the boron concentration is relatively stable. Therefore, deviations from the normal boron concentration band are generally small and the impact on a LOCA is minor. This is because boiling of the emergency core cooling water during the reflood phase of a LOCA concentrates the boron in the saturated liquid that remains in the core. In addition, the volume of the SIT is still available for injection. Since the boron requirements are based on the average boron concentration of the total volume of three SITs, the consequences are less severe than they would be if an SIT were not available for injection.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change extends the allowed outage time for an inoperable SIT. An extension in the allowed outage time for an inoperable component is not assumed to be an initiator of any evaluated accident. Therefore, extending the allowed outage time for an inoperable SIT does not involve a significant increase in the probability of an accident previously evaluated .

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ATTAC1'ENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

1. (continued)

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed mitigatory function of the remaining Operable SITs or other emergency core cooling system structures or components. Thus, the consequence of an accident occurring during the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowed outage time presently specified in the technical specifications is the same as the consequences for an accident occurring during an allowed outage time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to ensure prompt restoration of an inoperable SIT to re-establish compliance with the limiting condition for operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change increases the allowed outage time from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for an SIT whose boron concentration is not within limits, or when level or pressure of an SIT can not be verified. The margin of safety afforded by the SITs is in their ability to supply water to the reactor vessel during a LOCA and thus help ensure that the acceptance criteria, established by 10 CFR 50.46 for the emergency core cooling system, will be met. Since the SITs are passive components, their boron concentration, level and pressure are relatively stable and deviations in these parameters are generally small. Thus, the impact on a LOCA as a result of one of these parameters being outside their limit is minor since the contents of the inoperable SIT will still be available for injection in the event of a LOCA. In addition, an extension in the allowed outage time provides some period of time to restore an inoperable SIT to Operable status without initiating a plant shutdown which increases the chance of an undesired plant transient. As such, any reduction in a margin of safety resulting from an extended allowed outage time would most likely be offset from the benefits gained from restoring the inoperable SIT and precluding an unnecessary plant shutdown. Therefore,

this change does not involve a significant reduction in a margin of safety.

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

LESS RESTRICTIVE CHANGE L.3 CTS 3. 3. 2 contains a provision which allows one of the ECCS components required by CTS 3.3.1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements of LCO 3.0.3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss of ECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a"condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss of ECCS function will not occur. This

'" is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change will allow multiple components associated with the Emergency Core Cooling System (ECCS) to be simulataneously inoperable for a specified time provided the core cooling function is not lost. An extension in the allowed outage time for inoperable components is not assumed to be an initiator of any evaluated accident. Therefore, extending the allowed outage time for multiple component inoperabilities while preserving the overall function of the specified requirement does not involve a significant increase in the probability of an accident previously evaluated .

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ATTACHMENT 4

1. (continued)

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change only extends the time multiple ECCS components can be made inoperable without requiring a plant shutdown provided the functional capability of the system is maintained. Thus the consequences of an accident occurring during the allowed outage time presently specified in the technical specifications is the same as the consequences for an accident occurring during the proposed allowed outage time in the ITS.

Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to assure the approprite ECCS components are Operable to fulfill the core cooling function during a DBA.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change increases the time mutiple ECCS components can be made inoperable provided the core cooling function is maintained. The margin of safety afforded by the ECCS is related to the assurance that the acceptance criteria setforth in 10 CPR 50.46 will be met following a LOCA. Acceptable consequences following a LOCA can be achieved by a single train of ECCS components and 3 of 4 Safety Injection Tanks. The proposed change continues to ensure at least 100 % of the ECCS flow equivalent to a single Operable ECCS train, and 3 of 4 Safety Injection Tanks, are available. As such, removing more than one ECCS component at any given time does not affect affect the capability of the ECCS to perform its intended function. Any reduction in a margin of safety resulting from an extended allowed outage time would mostly likely be offset from the benefits gained from restoring the inoperable components to operation and thereby precluding an unnecessary shutdown. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

LESS RESTRICTIVE CHANGE L.4 CTS 4.6.4a and CTS 4.2, Table 4.2.2 item #10 contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion. Specifically, CTS 4.6.4a states that SIT motor operated isolation valve position is verified "by observing valve position indication and (the) valve itself." In addition, CTS 4.6.4a also requires a verification that the breakers for the SIT motor operated isolation valves are ldcked opened. The intent of verifying valve position and removing power from the valve operator is to ensure that the valve is in its correct position and that an active failure does not result in the undetected closure of a valve. Specifying that valve position be verified "by observing valve position indication and (the) valve itself", or stating that the "breaker must be locked opened" do not constitute requirements assumed in the safety analyses. Rather, they simply provide a method for assuring valve and breaker position. CTS 4.2 Table 4.2.2 item 10 contains a surveillance requirement to verify SIT pressure is less than the high alarm.

The purpose of this alarm is to alert the operator of an off normal condition (e.g., nitrogen inleakage or check valve back leakage). This requirement has been deleted since the safety analyses do not assume a maximum SIT pressure. Thus, it is not appropriate to contain a requirement whose function is strictly to alert the operator of an off normal condition when the Operability of the affected component is not impacted. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event.

The proposed change deletes details from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the

. probability or consequences of an accident previously evaluated.

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ATTAC1'ENT 4

2.

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.1, SAFETY INJECTION TANKS (SITs)

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

The proposed change deletes detail from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

I Margin of safety is determined by the design and qualification of the plant equipment, I the operation of the plant within analyzed limits, and the point at which protective or I mitigative actions are initiated. There are no design changes or equipment performance I parameter changes associated with this change. No setpoints are affected, and no *1 change is being proposed in the plant operational limits as a result of this change. The I proposed changes deletes details from the Technical Specifications. Removal of these I details is acceptable since this information is not directly pertinent to the actual I requirement and does not alter the intent of the requirement. Since these details are not I necessary to adequately describe the actual regulatory requirement, they can be moved I to licensee controlled document without a significant impact on safety. Therefore, the I proposed changes do not involve a significant reduction in a margin of safety. I

Palisades Nuclear Plant Page 7of7 04/09/99

ATTACHMENT 4

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING LESS RESTRICTIVE CHANGE L.1 CTS 3.3.2 requires the reactor to be placed in cold shutdown within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months when an ECCS component is not restored to Operable status within the allowed outage time. Proposed ITS 3.5.2 Required Action B.2 requires the PCS temperature to be reduce to < 325°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when an ECCS train is not restored to Operable status within the allowed outage time. Required Action B.2 places the plant in a condition in which two trains of ECCS are no longer necessary to mitigate the consequences of an accident. In MODE 3 with PCS temperature < 325°F, ECCS operational requirements are relaxed due to stable plant conditions and the reduced probability of a DBA. ECCS requirements in MODE 3 with PCS temperature < 325°F are addressed by proposed ITS 3.5.3, ECCS - Shutdown. Requiring the plant to be placed in a mode in which the LCO no longer applies is consistent with the philosophy of NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event.

The proposed change limits the condition in which the plant is placed when one ECCS train can not be restored to Operable to below the condition where two ECCS trains are credited in the safety analysis assuming a single failure. In MODE 3 with PCS temperature < 325 °F, ECCS operational requirements are relaxed due to stable plant conditions and the reduced probability of a DBA. As such, one ECCS train is required to mitigate a DBA without consideration for a single failure . By placing the plant in a condition which two ECCS trains are no longer necessary to mitigate the consequences of an accident there can be no impact on the probability or consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated .

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ATTAC1'ENT 4

2.

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the plant is placed in a condition in which two ECCS trains are not required to function during an accident assuming a single failure of one train. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Prompt and appropriate actions have been determined based on safety analysis assumptions to ensure the plant is placed in a conditions under which two ECCS trains are no longer required assuming the single failure of one train. Therefore, this change does not involve a significant reduction in the margin of safety.

LESS RESTRICTIVE CHANGE L.2

.CTS 3.3.2b, CTS 3.3.2c, CTS 3.3.2d, and CTS 3.3.2f require their respective inoperable ECCS component be restored to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Proposed ITS Condition A has revised the CTS to allow one or more ECCS train to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least 100% of the ECCS flow equivalent to a single Operable ECCS train is available.

Thus, the ITS allows a combination of HPSI and LPSI subsystems to be inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the flow assumed to be delivered by a single ECCS train is available. By stipulating a 100 % ECCS flow equivalent, proposed Condition A preserves the safety function of the ECCS system while allowing some period of time for correcting ECCS component inoperabilities. This allowance is acceptable because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. This change is supported by reliability analyses discussed in a NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975. This change is consistent with NUREG-1432 .

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ATTAC1'ENT 4

1.

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change extends the allowed outage time for one or more inoperable ECCS trains for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least 100 % of the ECCS flow equivalent to a single Operable ECCS train is available. An extension in the allowed outage time for an inoperable component is not assumed to be an initiator of any evaluated accident. Therefore, extending the allowed outage time for an inoperable ECCS train does not involve a significant increase in the probability of an accident previously evaluated.

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed mitigatory function of the remaining Operable ECCS components.

Thus, the consequence of an accident occurring during the allowed outage time presently specified in the technical specifications (LCO 3.0.3) is the same as the consequences for an accident occurring during an allowed outage time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to limit the amount of time the plant is allowed to operate with an inoperable ECCS train. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated .

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ATTACI'ENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING

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

The proposed change extends the allowed outage time for an inoperable ECCS component pump from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and the combination of an HPSI pump and LPSI pump from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provide at least 100% flow equivalent to a single ECCS train is available. The function of the ECCS is to provide core cooling and negative reactivity to ensure the core is protected following various accidents. By stipulating a 100 % ECCS flow equivalent as a condition to extend the inoperability of an ECCS train, the safety function of the ECCS system is preserved. This is because of the redundancy of trains and the diversity of subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its intended safety function. As such, the margin of safety associated with the extended allowed outage time remains unchanged. Therefore, this change does not involve a significant reduction in a margin of safety.

LESS RESTRICTIVE CHANGE L.3 CTS 3.3.2 contains a provision which allows one of the ECCS components required by CTS 3.3.1 to be made inoperable for a specified time provided the remaining components are Operable. Since the CTS does not provide an explicit action for multiple component inoperabilities, the plant would invoke the requirements of LCO 3.0.3 when two or more of the listed components are made inoperable. The purpose of CTS 3.3.2 is to ensure a loss of ECCS function does not occur by limiting the ECCS components that can be removed from service to only one component at any given time. The structure of the CTS is such that non-compliance with the LCO is addressed on a "component based" level. The ITS (and ISTS) is structured to address LCO non-compliance on a"condition based"level. As such, the ITS permits multiple component inoperabilities without a corresponding reduction in allowed outage time provided the functional requirements of the LCO are maintained. Although the actual requirements of the ITS are less restrictive than the CTS, the proposed change is acceptable since the ITS continues to ensure that a loss of ECCS function will not occur. This is assured by specifying "condition based" actions within a given specification that preserve the function of the LCO, and by evaluation performed in accordance with the Safety Function Determination Program for support system inoperabilities. Therefore, this change can be made without a significant risk to public health and safety. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 4 of 7

04/09/99

ATTAC1'ENT 4

1.

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change will allow multiple components associated with the Emergency Core Cooling System (ECCS) to be simulataneously inoperable for a specified time provided the core cooling function is not lost. An extension in the allowed outage time for inoperable components is not assumed to be an initiator of any evaluated accident. Therefore, extending the allowed outage time for multiple component inoperabilities while preserving the overall function of the specified requirement does not involve a significant increase in the probability of an accident previously evaluated.

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change only extends the time multiple ECCS components can be made inoperable without requiring a plant shutdown provided the functional capability of the system is maintained. Thus the consequences of an accident occurring during the allowed outage time presently specified in the technical specifications is the same as the consequences for an accident occurring during the proposed allowed outage time in the ITS.

Therefore, the proposed change does not involve a significant increase in the I consequences of an accident previously evaluated. I

I

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

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to assure the approprite ECCS components are Operable to fulfill the core cooling function during a DBA.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated .

Palisades Nuclear Plant Page 5of7 04/09/99

ATTAC1'ENT 4

3.

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING Does this change involve a significant reduction in a margin of safety?

The proposed change increases the time mutiple ECCS components can be made inoperable provided the core cooling function is maintained. The margin of safety afforded by the ECCS is related to the assurance that the acceptance criteria setforth in 10 CPR 50.46 will be met following a LOCA. Acceptable consequences following a LOCA can be achieved by a single train of ECCS components and 3 of 4 Safety Injection Tanks. The proposed change continues to ensure at least 100 % of the ECCS flow equivalent to a single Operable ECCS train, and 3 of 4 Safety Injection Tanks, are available. As such, removing more than one ECCS component at any given time does not affect affect the capability of the ECCS to perform its intended function. Any reduction in a margin of safety resulting from an extended allowed outage time would mostly likely be offset from the benefits gained from restoring the inoperable components to operation and thereby precluding an unnecessary shutdown. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

LESS RESTRICTIVE CHANGE L.4 CTS 4.6.la and CTS 4.6.3a contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion.

Specifically, CTS 4.6.lA contains the requirement for testing the Safety Injection System. The CTS states that satisfactory results of the test are indicated by "control board indication and visual observation." The intent of this test is to verify that components which receive an actuation signal actuate to their correct position. Specifying that valve position be verified "by "control board indication and visual observation." does not constitute a requirement assumed in the safety analyses. Rather, it simply provide a method for assuring valve position. CTS 4.6.3a states that "alternate manual starting (of the safety injection pumps and shutdown cooling pumps) between the control roo111 console and the local breaker shall be practiced in the test program." The ability to demonstrate the manual starting capability of the ECCS pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capability of meeting their intended safety function. Since the above details are not necessary to describe, or are not pertment to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 6of7 04/09/99

ATTACHMENT 4

1.

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.5.2, EMERGENCY CORE COOLING (ECCS) OPERATING Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event.

The proposed change deletes details from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change deletes detail from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes deletes details from the Technical Specifications. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a significant impact on safety. Therefore, the

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

Palisades Nuclear Plant Page 7of7 04/09/99

ECCS-Operat i ng B 3.5.2 BASES (J .

APPLICABLE On smaller breaks,~S pressure will stabilize at a value SAFETY ANALYSES dependent upon bre~ .size, heat load, and injection flow.

(continued) The smaller the break, the higher this equilibrium p~essure.

In all LOCA analyses, injection flow is not credited until f @:S pressure drops below the shutoff head of the HPSI pumps.

de.u~----'

o."'4 * :,.,.Moo~ . ..C ftS fc.'""""~+.Jr.._ ~ 3~5 ° f 1 LCO In MODES lJ 2, and (Jtwith (presslJli zer pri'ssure i/'1700 osA a) two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming there is a single failure affecting either train .

Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.

n an n i e ,

700 ia ,in CCS train consists of an HPSI subsyste~ and a.

~ ensure the availability L SI su syste~atid i/chartjljng pulilb>. ..J,.., o.d,d..1-w, Ee.LS '

J'ach~train includes the piping, instruments, and controls to Q..f._an OPERABLE flow path capable of sl~wT .....J.1.k.ing__$Jt~.U!?.~tlrQ!ll---1b.§CM on an S~ and automatically transferring suction to the containment sump upon a Jecirculation ,.fctuation filignal (RAS).

During an event requiring ECCS actuation, a flow path is~SiluJ'f provided to ensure an abundant supply of water from the to the~S, via the HPSI and LPSI pumps and their respect ve supply headers, to each of the four cold leg injection

..1.NS£t:r I _____n_oz_z_i_e_s_.f In the long term, this flow path may be switched (continued)

CEOG STS B 3.5-13 Rev 1, 04/07/95 Revised 04/09/99

INSERT During the recirculation phase, a flow path is provided from the containment sump to the PCS via the HPSI pumps. For worst case conditions, the containment building water level alone is not sufficient to assure adequate Net Positive Suction Head (NPSH) for the HPSI pumps.

Therefore, to obtain adequate NPSH, a portion of the Containment Spray (CS) pump discharge flow is diverted from downstream of the shutdown cooling heat exchangers to the suction of the HPSI pumps following recirculation during a large break LOCA. In this configuration, the CS pumps and shutdown cooling heat exchangers provide a support function for HPSI flow path OPERABILITY. The OPERABILITY requirements for the CS pumps and shutdown cooling heat exchangers are addressed in LCO 3.6.6, "Containment Cooling Systems."

Support system OPERABILITY is addressed by LCO 3.0.6.

B 3.5-13 Revised 04/09/99

ECCS-Operat i ng B 3.5.2 BASES LCO to take supply from the contain ent sump and to supply (continued) part of flow to the RCS hot le s via the shutdown cooling DC) suction nozzles. Te charging pump flow p takes s tion from the RWT ands pplies the RCS via the

...,. normal harging lines .

The flow path for each train must maintain its designed independence to ensure that no single~failure can disable both ECCS trains. o.etiv~

f

<JV APPLICABILITY In MODES I and 2, and in MODE 3 with ~SfP(essure-i fem~~ ~ ~.2.S oF 11m~in9esign

~l~O ~ the ECCS OPERABILITY requirements for the Basis Accident (OBA) large break LOCA are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The HPSI pump performance is based on the small break LOCA, which establishes the um er nee curve and has less de endence on owe . he charging p p

\ er ance re u1r en s e based o a sm The requirements of MODE 21 andl3, w1th;&S res Ci}f70Cf'IPJ!]l, are bounded by the MODE 1 analysis.

. ~b~ p The ~CCS functional requirements of MODE 3, with <<cs

$re(syre </]700 Ws1), and MOOE 4 are described in LCO 3.5.3,

ECCS-Shutdown.*

~1' In HODES 5 and 6,~conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled,"

and LCO 3.4.8, *Res Loops-MODE 5, Loops Not Filled."

MOOE 6 core cooling requirements are addressed by LCO 3.9.4, "Shutdown Cooling {SOC) and Coolant Circulation-High Water Level," and LCO 3.9.S, "Shutdown Cooling (SOC) and Coolant Circulation-Low Water Level."

ACTIONS

{continued)

CEOG STS B 3.5-14 Rev 1, 04/07/95 Revised 04/09/99

ECCS-Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.5 REQUIREMENTS (continued)

SR 3.5.2j' SR 3:5.2~and SR 3.5.2.1?

These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated

~---.SIB; caKd jtl 2il RAS, that each ECCS pump starts on receipt of

~ an actUal or simulated s(IJ;, and that the LPSI pumps stop on rece1p o an actua or simulatedARAS. This Surveillance is

~ocs---,r=e'""'q""'uTC1r::=earorva 1ves that are l ocked , seal ed , or otherwise secured in the required position under administrative controls. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also accepta ed on consideration of the desi n reliability t_Xand operating ex erienc . e e ui m . The c a ion o ic is es e as par o the Engineered Safety Feature

st (ESlt&S) testing, and equipment performance is mon to ed as part of the Inservice Testing Program.

SB 3.5.2.D va ves in proper ECCS (continued)

CEOG STS B 3. 5-18 Rev 1, 04/07/95

Revised 04/09/99

SECTION 3.5 INSERT The HPSI Hot Leg Injection motor operated valves and the LPSI loop injection valves have position switches which are set at other than the full open position. This surveillance verifies that these position switches are set properly.

The HPSI Hot leg injection valves are manually opened during the post-LOCA long term cooling phase to admit HPSI injection flow to the PCS hot leg. The open position limit switch on each HPSI hot leg isolation valves is set to establish a predetermined flow split between the HPSI injection entering the PCS hot leg and cold legs. ,

The open position limit switch on each LPSI loop injection valve is set to establish the maximum possible flow through that valve. The design of these valves is such that excessive turbulence is developed in the valve body when the valve disk is at the full open position. Stopping the valve travel at slightly less than full open reduces the turbulence and results in increased flow.

Verifying that the position stops are properly set ensures that a single low pressure safety .

injection subsystem is capable of delivering the flow rate required in the safety analysis .

~..

B3.5-18 Rev;sed 04/09/99

ECCS-Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

Rc.s.: R:s B 3.5.3 ECCS-Shutdown (j) K\JJ r -. SIRW1 HPs1 : LP.s-t BASES BACKGROUND APPLICABLE The Applicable Safety Analyses section of Bases. 3.5.2 is SAFETY ANALYSES applicable to these Bases.

Due to the stable conditions associated with operation in MODE 4, and the reduced probability of a Design Basis Accident (OBA), the ECCS operational requirements are reduced. Included in these reductions is that certain automatic safety injection (SI) actuation signals are not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a OBA.

Only one train of ECCS is required for MODE 4. Protection against single failures is not relied on for this MODE of operation.

~JC~~itdown satisfies Criterion 3 of tt!je NR¢ Polity)

(Sta e . / D C.~f So.~k Cc. 'Cc)

t. fcs -ff.,,., Pf. *a-tvr<. < ~ z..s o f c.u1 d "' mcd <. 4 LCO In MODE 3 with (p.r.eriyrizet pressij'r.e < l/i)Q o~a>. an EC~S irtl.;"

(~~~is composed of a singleL~SI subsystem. Eaclt-<SPSI includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the~ and transferring suet ion to the containment sump. *

~StlW( <

(continued)

CEOG STS B 3.5-20 Rev 1, 04/07/95

Revised 04/09/99

SECTION 3.5 INSERT 1 In Mode 3 with PCS temperature < 3 25 °F and in Mode 4 the normal compliment of ECCS components is reduced from that which is available during operations above Mode 3 with PCS temperature ~ 325 °F. The acceptability for the reduced ECCS operational requirements is based on engineering judgement rather than specific analysis and considers such factors as the reduced probability that a LOCA will occur, and the reduced energy stored in the fuel. The reduction in ECCS operational requirements include:

1) Isolation of the Safety Injection Tanks (SITs) since PCS pressure is expected to be reduced below the SIT injection pressure,

2) Reliance on manual safety injection initiation since the automatic Safety Injection Signal (SIS) is not required by the technical specifications below 300°F,

3) Rending the High Pressure Safety Injection (HPSI) pumps incapable of injecting into the PCS. The HPSI pumps are rendered incapable of injecting into the PCS in accordance with the requirements ofLCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System". This action assures that a single mass addition event initiated at a pressure within the limits of LCO 3 .4 .12 cannot cause the PCS pressure to exceed the 10 CFR 50 Appendix G limit.

At a PCS temperature of325°F the maximum allowed PCS pressure corresponds to the LTOP setpoint limit which is approximately 800 psia. Below 800 psia postulated piping flaws of critical size are considered unlikely since normal operation at 2060 psia serves as a proof test against ruptures. In addition, since the reactor has been shutdown for a period of time, the decay heat and sensible heat levels are greatly reduced from the full power case.

Although a pipe break in the PCS pressure boundary is considered unlikely, break sizes larger and smaller than approximately O. lft2 are considered separately when analyzing ECCS response.

For breaks larger than approximately O. lft:2, the event is characterized by a very rapid depressurization of the PCS to near the containment pressure. Due to the reduced temperature and pressure of the PCS, the time to complete blowdown is extended from that assumed in the full power case. During this time, the fuel is cooled by the flow through the core towards the break. Automatic safety injection actuation is not assumed to occur since the pressurizer pressure SIS maybe bypassed below 1700 psig. Therefore, operator action is relied upon to initiate ECCS flow. Indication that would alert the operator that a LOCA had occurred include; a loss of pressurizer level, rapid decrease in PCS pressure, increase in containment pressure, and containment high radiation alarm. Since the saturation pressure for 325 °F is approximately

100 psia, the LPSI pumps are capable of providing the required heat removal function. When the OPERABLE LPSI pump is being used to fulfill the shutdown cooling function, the PCS pressure is< 300 psia. As such, the rate of PCS blowdown is reduced providing some time to manually realign the OPERABLE LPSI pump to the ECCS mode of operation.

B 3.5-20 Revised 04/09/99

SECTION 3.5 INSERT 1 For breaks smaller than approximately O. Ift:2, the event is characterized by a slow depressurization of the PCS and a relatively long time for the PCS level to drop below the tops of the hot legs. In MODE 3 with PCS temperature< 325 °F and in the upper range of MODE 4 before shutdown cooling is established, the spectrum of smaller break sizes are more limiting than larger breaks in terms ofECCS performance since the PCS could stay above the shutoff head of the LPSI pumps. For these break sizes, sufficient time, well in excess of recommended 10 minutes attributed for manual operator action, is available to either initiate once through cooling using the PORVs, or by re-establishing HPSI pump injection capability. In either case, the core remains covered and the criteria of 10 CFR 50.46 preserved.

B 3.5-20 Revised 04/09/99

ENCLOSURE 4 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO APRIL 12, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.8, ELECTRICAL POWER SYSTEMS

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO APRIL 12, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.8, ELECTRICAL POWER SYSTEMS Page Change Instructions Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.

REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT#

ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL ITS 3.8.4-1 ITS 3.8.4-1 04/30/99 1 &2 ITS 3.8.4-2 ITS 3.8.4-2 04/30/99 1 &2 ITS 3.8.4-3 ITS 3.8.4-3 04/30/99 ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL ITS 8 3.8.4-3 ITS 8 3.8.4-3 04/30/99 1 ITS 8 3.8.4-4 ITS 8 3.8.4-4 04/30/99 2 ITS 8 3.8.4-5 ITS 8 3.8.4-5 04/30/99 1 &2 ITS 8 3.8.4-6 ITS 8 3.8.4-6 04/30/99 ITS 8 3.8.4-7 ITS 8 3.8.4-7 04/30/99 ITS 8 3.8.4-8 ITS 8 3.8.4-8 04/30/99 ITS 8 3.8.4-9 ITS B 3.8.4-9 04/30/99 ITS 8 3.8.4-10 ITS 8 3.8.4-10 04/30/99 ITS B 3.8.4-11 ------------- 04/30/99 ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL CTS 3.8.4, pg 3-45 CTS 3.8.4, pg 3-45 04/30/99 1 &2 DOC 3.8.4, pg 2 of 3 DOC 3.8.4, pg 2 of 3 04/30/99 1 &2 ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL NSHC 3.8.4, pg 1 of 4 NSHC 3.8.4, pg 1 of 4 04/30/99 1 &2 NSHC 3.8.4, pg 2 of 4 NSHC 3.8.4, pg 2 of 4 04/30/99 1 &2 ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG 3.8-25 NUREG 3.8-25 04/30/99 1 &2 NUREG 3.8-25 insert NUREG 3.8-25 insert 04/30/99 1 &2 ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL JFD 3.8.4, pg 2 of 2 JFD 3.8.4, pg 2 of 2 04/30/99 1 &2 ATTACHMENT 7 TO ITS CONVERSION SUBMITTAL CTS 3.8.4, pg B 3.7.4-3 CTS 3.8.4, pg B 3.7.4-3 04/30/99 1 &2 CTS 3.8~4, CTS 3.8.4, pg B 3.7.4-3 insert pg B 3.7.4-3 insert 04/30/99 1 &2 CTS 3.8.4, pg 8 3.7.4-4 CTS 3.8.4, pg 8 3.7.4.4 04/30/99 1 &2 CTS 3.8.4, pg B 3.7.4-4 insert ---------------- 04/30/99 1 &2 ATTACHMENT 8 TO ITS CONVERSION SUBMITTAL No page changes.

ATTACHMENT 9 TO ITS CONVERSION SUBMITTAL No page changes.

1

DC Sources - Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources - Operating LCO 3.8.4 The Left Train and Right Train DC electrical power sources shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required DC A.1 Verify functional 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power cross-connected source battery charger battery charger is inoperable. connected supplying power to the affected DC train.

AND A.2 Restore required DC 7 days electrical power source battery charger to OPERABLE status.

B. One required DC B.1 Verify OPERABLE 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power directly connected source battery and functional inoperable. cross-connected battery chargers are connected supplying power to the affected DC train.

AND B.2 Restore required DC 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months electrical power source battery to

Palisades Nuclear Plant OPERABLE status.

3.8.4-1 Amendment No. 04/30/99

DC Sources - Operating 3.8.4

ACTIONS C.

CONDITION Required Action and C.l REQUIRED ACTION Be in MODE 3.

COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is ~ 125 V 7 days on float charge .

SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.

Verify battery connection resistance is 92 days s 50 µohm for inter-cell connections, s 360 µohm for inter-rack connections, and s 360 µohm for inter-tier connections.

SR 3.8.4.3 Inspect battery cells, cell plates, and 12 months racks for visual indication of physical damage or abnormal deterioration that could degrade battery performance.

SR 3.8.4.4 Remove visible terminal corrosion and 12 months verify battery cell to cell and terminal connections are coated with anti-corrosion material .

Palisades Nuclear Plant 3.8.4-2 Amendment No. 04/30/99

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.5 Verify battery connection resistance is 12 months

~ 50 µohm for inter-cell connections,

~ 360 µohm for inter-rack connections, and

~ 360 µohm for inter-tier connections.

SR 3.8.4.6 Verify each required battery charger 18 months supplies 2 180 amps at 2 125 V for 2 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SR 3.8.4.7 -------------------NOTES-------------------

1. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7.

2. This Surveillance shall not be

performed in MODE l, 2, 3, or 4.

Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the design 18 months duty cycle when subjected to a battery service test .

Palisades Nuclear Plant 3.8.4-3 Amendment No. 04/30/99

DC Sources - Operating B 3.8.4

BASES LCD The LCD requires chargers ED-15 and ED-16 because those (continued) chargers are powered by the AC power distribution system and DG associated with the battery they supply. If only the cross connected chargers were available, and a loss of off-site power should occur concurrently with the loss of one DG, both safeguards trains would eventually become disabled. One train would be disabled by the lack of AC motive power; the other would become disabled when the battery, whose only OPERABLE charger is fed by the failed DG, became depleted.

The required chargers, ED-15 and ED-16, must be OPERABLE, but need not actually be in service because the probability of a concurrent loss of offsite power with loss of one DG is low, and battery charging current is not needed immediately after an accident.

APPLICABILITY The DC sources are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that redundant sources of DC power are available to support engineered ~afeguards equipment and plant instrumentation in the event of an accident or transient. The DC sources also support the equipment and instrumentation necessary for power operation, plant heatups and cooldowns, and shutdown operation.

The DC source requirements for MODES 5 and 6, and during movement of irradiated fuel assemblies are addressed in LCD 3.8.5, "DC Sources - Shutdown."

ACTIONS A.I and A.2 With one of the required chargers (ED-15 or ED-16) inoperable, the cross connected charger must be placed in-service within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , if it is not already in service, to maintain the battery in OPERABLE status. If the cross connect charger is not placed in service within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , Condition C would be entered.

Additionally for the cross-connected charger to be considered "functional," the cross-connected charger must have been surveilled and satisfied the same performance test required for the directly connected charger (i.e.,

SR 3.8.4.6) within the required Frequency.

Palisades Nuclear Plant B 3.8.4-3 04/30/99

DC Sources - Operating B 3.8.4

BASES ACTIONS A.1 and A.2 (continued)

In order to limit the ti~e when the DC source is not capable of continuously meeting the single failure criterion, the required charger must be restored to OPERABLE status within 7 days.

The 7 day completion time was chosen to allow trouble shooting, location of parts, and repair.

B.1 and B.2 With one battery inoperable, the associated DC system cannot meet its design. It lacks both the surge capacity and the independence from AC power sources which the battery provides if offsite power is lost. Placing the second battery charger in service provides two benefits:

1) restoration of the capacity to supply a sudden DC power demand, and 2) restoration of adequate DC power in the affected train as soon as either AC power distribution system is re-energized following a loss of offsite power.

If the cross connect charger is not placed in service within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , Condition C would be entered. Additionally for the cross connected charger to be considered "functional," the cross connected charger must have been surveilled and satisfied the same performance test required for the directly connected charger (i.e., SR 3.8.4.6) within the required Frequency.

In order to restore the DC source to its design capability, the battery must be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is a feature of the original Palisades licensing basis and reflects the availability to provide two trains of DC power from either AC distribution system. Furthermore, it provides a reasonable time to assess plant status as a function of the inoperable DC electrical power source and, if the battery is not restored to OPERABLE status, to prepare to effect an orderly and safe plant shutdown .

Palisades Nuclear Plant B 3.8.4-4 04/30/99

DC Sources - Operating B 3.8.4

BASES ACTIONS C.1 and C.2 (continued)

If the inoperable DC electrical power source cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 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.

SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous current required to overcome the internal losses of a battery and maintain the battery in a fully charged state. The specified voltage is the nominal rating of the battery. At that terminal voltage, the battery has sufficient charge to provide the analyzed capacity for either accident loading or station blackout loading. The 7 day Frequency is consistent with manufacturer and IEEE-450 (Ref. 4) recommendations.

SR 3.8.4.2 Visual inspection to detect corrosion of the battery terminals and connectors, or measurement of the resistance of each inter-cell and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.

The specified limits of~ 50 µohm for inter-cell connections and terminal connections, and ~ 360 µohms for inter-tier and inter-rack connections are in accordance with the manufacturers recommendations.

Palisades Nuclear Plant B 3.8.4-5 04/30/99

DC Sour~es - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.2 (continued)

REQUIREMENTS The 50 µohm value is based on the minimum battery design voltage. Battery sizing calculations show the first minute load on- the ED-02 battery as the load that determines battery size, hence, battery voltage will be at its lowest value while the battery supplies this current. Calculations also show that at a minimum temperature and end of life (80% battery performance), battery voltage during this first minute load will be about 1.815 V per cell, assuming nominal connection resistance. But if all the connections were at the ceiling value of 50 µohms, the battery manufacturer indicates that the additional voltage drop would result in a battery voltage of about 1.79 V per cell, which is still above the minimum design voltage (Ref. 5).

The 360 µohm value is based on 120% of the nominal cumulative resistance of the components which make up the connections: resistance of the connecting cable, and for each end of the cable, the battery post to cable lug connection, the cable lug itself, and the lug to cable connection.

The resistance values determined during initial battery installation are recorded with the battery replacement specifications, FES 95-206-ED-01 and FES 95-206-ED-02.

The Surveillance Frequency for these inspections, which can detect conditions that can cause power losses due to resistance heating, is 92 days. This Frequency is considered acceptable based on operating experience related to detecting corrosion trends.

SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance. The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function).

Palisades Nuclear Plant B 3.8.4-6 04/30/99

DC Sources - Operating B 3.8.4

BASES SURVEILLANCE SR 3.8.4.3 (continued)

REQUIREMENTS The 12 month Frequency for this SR is consistent with IEEE-450 (Ref. 4), which recommends detailed visual inspection of cell condition and rack integrity on a yearly basis.

SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of inter-cell and terminal connections provide an indication of physical damage or abnormal deterioration that could indicate degraded battery condition. The anticorrosion material is used to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection. The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR provided visible corrosion is removed during performance of SR 3.8.4.4.

The specified limits for connection resistance are discussed in the Bases for SR 3.8.4.2.

The Surveillance Frequencies of 12 months is consistent with IEEE-450 (Ref. 4), which recommends cell to cell and terminal connection resistance measurement on a yearly basis.

SR 3.8.4.6 This SR requires that each required battery charger be capable of supplying 180 amps at 125 V for ~ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

These requirements are based on the design capacity of the chargers. The chargers are rated at 200 amps; the specified 180 amps provides margin between the charger rating and the test requirement .

Palisades Nuclear Plant B 3.8.4-7 04/30/99

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.6 (continued)

REQUIREMENTS The specified Frequency requires each required battery charger to be tested each 18 months. The Surveillance Frequency is acceptable, given the other administrative controls existing to ensure adequate charger performance during these 18 month intervals. In addition, this Frequency is intended to be consistent with expected fuel cycle lengths.

SR 3.8.4.7 A battery service test is a special test of battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length should correspond to the*

design duty cycle requirements as specified in FSAR Chapter 8 (Ref. 2).

The Surveillance Frequency of 18 months is consistent with the recommendations of RG 1.32 (Ref. 6) and RG 1.129 (Ref. 7), which state that the battery service test should be performed during refueling operations, or at some other outage, with intervals between tests not to exceed 18 months ..

Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8 while satisfying the requirements of SR 3.8.4.7 at the same time.

The reason for the restriction that the plant be outside of MODES 1, 2, 3, and 4 is that performing the Surveillince requires disconnecting the battery from the DC distribution buses and connecting it to a test load resistor bank. This action makes the battery inoperable and completely unavailable for use.

Palisades Nuclear Plant B 3.8.4-8 04/30/99

DC Sources - Operating B 3.8.4

BASES SURVEILLANCE SR 3.8.4.8 REQUIREMENTS (continued) A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as 11 found" condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.

The modified performance discharge test is a simulated duty cycle consisting of just two rates; the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelop the duty cycle of the service test. Since the ampere-hours removed by a rated one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

A modified performance discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test.

Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8 while satisfying the requirements of SR 3.8.4.7 at the same time.

The acceptance criteria for this Surveillance are consistent with the recommendations of IEEE-450 (Ref. 4) and IEEE-485 (Ref. 3). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.

Palisades Nuclear Plant B 3.8.4-9 04/30/99

DC Sources - Operating 8 3.8.4

BASES SURVEILLANCE SR 3.8.4.8 (continued)

REQUIREMENTS The Surveillance Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is

< 100% of the manufacturer*s rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity ~ 100% of the manufacturer*s rating. Degradation is indicated, according to IEEE-450 (Ref. 4), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is ~ 10% below the manufacturer*s rating. These Frequencies are consistent with the recommendations in IEEE-450 (Ref. 4).

The reason for the restriction that the plant be outside of MODES 1, 2, 3, and 4 is that performing the Surveillance requires disconnecting the battery from the DC distribution buses and connecting it to a test load resistor bank. This action makes the battery inoperable and completely unavailable for use.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17

2. FSAR, Chapter 8

3. IEEE-485-1983, June 1983

4. IEEE-450-1995

5. Letter; Graham Walker, C&D Charter Power Systems, Inc to John Slinkard, Consumers Power Company, 12 July 1996

6. Regulatory Guide 1.32, February 1977

7. Regulatory Guide 1.129, December 1974 Palisades Nuclear Plant 8 3.8.4-10 04/30/99

3.<i.+..1 ELECTRICAL PQWER SYSTEMS 3.1.~ 3,7,4 DC Soyrces - Operatjnq

a. D-01 a"jl"tharger ED-15, and

b. ED-02~d Charger ED-16.

Appljcability (~ l"l. 3 a.~ ~

plan*~:~v~Q*H~ ~. 0..:},J 1

Specification 3.7.4 applies when the .

Action 3.7.4.A inoperable: ./~

battery in*'

2. Restore the required charger to OPERABLE status; within 7 days.

3.7.4.B With one battery inoperable: ~

1. Place o charger~n service for the ~cted battery;

]rnme iatelY. and t!:..3J

2. Restore the battery to OPERABLE status; within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(Md C.. 3. 7 .4.C If any action required by 3.7.4.A or 3.7.4.B 1s not met and the associated completion time has expired: ~

1. The reactor shall be placed in HOT SHUTDOWN; within~':.":~-~

2. The reactor shall be placed in COLD SHUTDOWN; within ~rs.~

(!9

/i .

I ' ' '

3-45

Amendment No.

Revised 04/30/99

ATTACI'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.8.4, DC SOURCES- OPERATING LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

LA.1 CTS 3.7.4 and 4.7.4.6 detail the plant equipment designators for the DC electrical power trains. ITS 3.8.4 simply requires the operability of the DC electrical power trains, and relocates the details of what comprises the trains to the Bases and FSAR.

Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to a licensee controlled document without a significant impact on safety. Placing these details in the Bases of ITS 3. 8 .4 and the FSAR provides adequate assurance that they will be maintained. The Bases are controlled by the Bases Control Process in Chapter 5 of the proposed Technical Specifications, and the FSAR is controlled by 10 CFR 50.59. This change is consistent with NUREG-1432.

LESS RESTRICTIVE CHANGES (L)

L.1 CTS 3. 7.4 specifies the requirements for DC source when the plant is above Cold Shutdown. CTS 3. 7.4 Actions A. l require the cross-connected charger(s) for the affected battery be placed in service immediately whenever one required charger is inoperable. CTS 3.7.4 Action B.1 requires both chargers for the affected battery be placed in service immediately whenever one battery is inoperable. The Completion Times for ITS Actions A. l and B. l have been extended to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months since these conditions are ones where DC power continues to remain available to the bus, albeit with inoperable sources. Selection of the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is More Restrictive than that previously approved for the more degraded condition of one de-energized DC electrical distribution train allowed by CTS 3.7.9.c. With one DC electrical source inoperable, the remaining DC electrical train is capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe condition.

Since ITS 3.8.4 represents a configuration of the DC electrical system that is less severe and presents less risk for a loss of function than that allowed by CTS 3.7.9.c (proposed ITS 3.8.9), an extension of the Completion Times to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months can be made without a significant risk to public health and safety. An extension of these Completion Times for "DC Source-Operating" is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 2 of 3 04/30/99 L

ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.8.4, DC SOURCES- OPERATING LESS RESTRICTIVE CHANGE L.1 CTS 3. 7. 4 Actions A .1 and B .1 both have requirements for connection of battery charger(s)

"immediately." These conditions are ones where* DC power continues to remain available to the bus, albeit with inoperable sources. In the event the entire DC bus is de-energized, CTS 3.7.9.C.2 allows 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to restore power prior to commencing a required plant shutdown.

Therefore, the ITS actions to connect additional battery charger(s) for degraded (but still powered) DC trains, are being proposed with a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allowed completion time. Without this change, the CTS actions result in the plant being in a shutdown action immediately upon discovery of an inoperable battery or charger (1,mtil the opposite train spare charger is connected).

Therefore, this change allows sufficient time to establish connection of the opposite train spare charger which may avoid a potentially unnecessary plant shutdown transient. The time allowed is More Restrictive than that previously approved for the more degraded condition of the de-energized train. As such, this change will not result in a significant impact on safety.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The actions for establishing the connection of an additional spare charger when operating with an inoperable battery or inoperable battery charger are proposed to be extended from "immediately" to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . These times are More Restrictive than the time approved for a de-energized DC train. This cha:r~ge will not result in a significant increase in the probability of an accident previously evaluated because an inoperable DC source is not the initiator of any analyzed event.

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The analyses which involve crediting the function of DC Sources continue to be supported by the remaining Operable sources. This change does not affect the performance of the minimum credited equipment. As a result, no analysis assumptions are violated. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.

Palisades Nuclear Plant Page 1of4 04/30/99

2.

NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.8.4, DC SOURCES - OPERATING Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or those procedures relied upon to mitigate a design basis event. Relaxing the allowed restoration time for an inoperable DC source does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure modes are being introduced. In addition, the change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. Sufficient DC sources remain available to support mitigating an analyzed event. The proposed change introduces action times for a degraded train of DC sources which are More Restrictive than the time approved for a de-energized DC train. Since the degradation continues to provide at least (if not more than) the equivalent safety function of the de-energized train, there is not reduction in the margin of safety from that previously approved.

LESS RESTRICTIVE CHANGE L.2 CTS 4.7.4.6 requires periodic testing of cross-connected battery chargers (ED-17 and ED-18),

however, the Specification does not require operability of these chargers. Since there is no specification that requires their operability, nor any specification that requires actions if the cross-connected chargers are inoperable (including not properly tested), the CTS surveillance is not appropriate for inclusion in the ITS. These cross-connected chargers are only required when a required battery or charger is inoperable. At that time, actions require connection of the cross-connected battery charger. In the event that the cross-connected charger(s) are used to satisfy this action, only then do they need to have been appropriately surveillanced .

Palisades Nuclear Plant Page 2 of 4 04/30/99

DC Sources-Operating 3.8.4

3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC LCO 3.8.4 Sources-O~C8J l

~

R.*a ~~

The rain~ and rain.l( DC electrical power ~ sha 11 be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

\

/N~Cl'2-r A. One DC ectrical A.I l Restore DC electrical power ubsystera power ~ubsystem to inop able OPERABLE status.

Required Action and Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met *

Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is 7 days

~ V on float charge.

0 (continued)

CEOG STS *3.8-25 Rev 1, 04/07/95

Revised

. 04/30/99

SECTION 3.8

ACTIONS CONDITION INSERT REQUIRED ACTION COMPLETION TIME A. One required DC electrical A. l Verify functional cross- 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months power source battery connected battery charger charger inoperable. is connected supplying power to the affected DC.

train.

ANll ,

A.2 Restore required DC 7 days electrical power source battery charger to OPERABLE status.

. B. One required DC electrical B.l Verify OPERABLE 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months power source battery directly connected and inoperable. functional cross-connected battery chargers are connected supplying power to the affected DC train.

ANll B.2 Restore required DC 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months '

electrical power source battery to OPERABLE status .

3.8-25 Revised 04/30/99

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.8.4, DC SOURCES - OPERATING

5. This change reflects the current licensing basis/technical specifications as modified by DOC L. l. ITS 3. 8 .4 has been developed to preserve the operational flexibility of the CTS while maintaining a parallel structure to the ISTS. The design of the Palisade plant DC electrical power system is such that each station battery has two associated battery chargers, one powered by the associated AC power distribution system (the directly connected charger), and one powered from the opposite AC power distribution system (the cross connected charger). CTS 3.7.4 provides the actions necessary to cope with a degraded DC electrical power system which results from an inoperable charger or an inoperable battery using the design features (i.e., cross connect capabilities) of the system. These same actions have been incorporated in ITS 3.8.4.

6. TSTF-8, Rev. 2 deleted the 3.8 SR Notes "However credit may be taken ... " and addressed that issue globally in the 3.0 Bases. This presentation is consistent with NUREG-1432.

7. Incorporate TSTF-38 into SR 3.8.4.3 to specify the surveillance is to identify conditions which could degrade battery performance .

Palisades Nuclear Plant _ Page 2 of 2 04/30/99

DC Sources - Operating B 3.7.4 and 4.7.4

BASES The required c argers, ED-15 and ED-16, must be OPERABLE, but need not actually be in service because the probability of a concurrent loss of offsite power oss of one DG is 10~

1 ~ry charging current is not needed immediately after an accident, ~- ~~ standha l1a1 ge1 s ffla:y Ile pl aced Wi Sif'Vt(i.Q. fjtlteltly. : dV\ti(

APPLICABILITY The DC sources are required to be OPERABLE to ensure that redundant sources of DC power are available to support engineered safeguards equipment and plant instrumentation in the event of an accident or transie~t.

The DC sources also support the equipment and instrumentation necessary for power operation,* plant heatups and cooldowns, and shutdown operation.

The DC source requirements for ~ObQ ~£~.s Rf;'t~~~~ and during movement of irradiated fuel assemblies are addressed in LCO 3>-..:.5, "DC Sources

- Shutdown." ~

ACTIONS I tJ 5£12..T ./

- ~A.I and A.2 With one of the required char~ers£iED-15 or ED-16) inoperable the cross connected charger must be \irn¢ dja 10 placed in serv1c , if it is not already in service, to maintain the battery in OPERABLE status. In order to limit the time when the DC source is not capable of continuously meeting the single failure criterion, the required charger must be restored to OPERABLE status within 7 days.

The 7 day completion time was chosen to allow trouble shooting, location of parts, and repair.

0 ~B.1 and B.2 With one battery inoperable, the associated DC system cannot meet its design.

It lacks both the surge capacity and the independence from AC power sources which the battery provides if offsite power is lost. Placing the second battery charger in service provides two benefits: 1) restoration of the capacity to supply a sudden DC power demand, and 2) restoration of adequate DC power in the affected train as soon ~s either AC power distribution system is re-energized following a loss of af~wer. /~/!:"fL.."r ~ )

PALISADES B 3.7.4°-3 Amendment No.

Revised 04/30/99 J

SECTION 3.8

INSERT I If the cross connect charger is not placed in service within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , Condition C would be entered. Additionally for the cross-connected charger to be considered "functional," the cross-connected charger must have been surveilled and satisfied .the same performance test required for the directly connected charger (i.e., SR 3.8.4.6) within the required Frequency.

INSERT 2 If the cross connect charger is not placed in service within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , Condition C would be entered. Additionally for the cross-connected charger to be considered "functional," the cross-connected charger must have been surveilled and satisfied the same performance test required for the directly connected charger (i.e., SR 3.8.4.6) within the required Frequency .

B 3.7.4-3 Revised 04/30/99

I DC Sources - Operating B 3.7.4 and 4.7.4 BASES In order to restore the DC source to its design capability, the battery must be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is a feature of the original Palisades licensing basis and reflects the availability to provide two trains of DC power from either AC distribution system. Furthermore, it provides a reasonable time to assess plant status as a function of the inoperable DC electrical power source and, if the battery is not restored to OPERABLE status, to prepare to effect an orderly and safe plant shutdown.

3.Z.4.C.1 111d e.e ... @*~- O.\o'\.

If the inoperable DC electrical power source cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to an operating condition in which the LCO does not apply. To achieve this status; the plant must be brought to at least within ours an o

~~RA'!Ht--withi hours. The allowed Completion Times are reasonable, based on operating experience, to reach the equired plant conditions from full power con l ions in an orderly manner and without challenging plant systems.

f(loD ~

SU LLANCE REQUIREMENTS J,g 4 (Fl eat velta~e eReek)-Q...

Verifying battery terminal voltage while on float charge helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger i~ supplying the continuous current required to overcome the internal losses *of a battery and maintain the battery in a fully charged. state. The specified voltage is the~nominal rating of the battery. At that terminal voltage, the battery has sufficient charge to provide the analyzed capacity for either accident loading or station blackout loading. The 7 day Frequency

s .S s; stent with manufacturer and IEEE,450( recommend at 1ons.1e..~ \ ' \

~ ~_/

3 4

. 4 (TerRli Ral a"d co"nectei* eo"eiti eR eReek) Q..

Visual inspection to detect corrosion of the battery terminals and connectors, or measurement of the resistance of each inter-cell and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.

The specified limits of s 50 µohm for inter-cell connections and terminal connections, and s 360 µohms for inter-tier and inter-rack connections are in accordance with the manufacturers recommendations.

PALISADES B 3.7.4-4 Amendment No.

Rev;sed 04/30/99

ML18068A585

Text

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