Text

Proposed Tech Specs Section 3.7,converting to ITS
	ML18066A443
Person / Time
Site:	Palisades
Issue date:	03/30/1999
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18066A442	List: ML18066A441; ML18066A443;
References
	NUDOCS 9904060084
	Download: ML18066A443 (278)
	v • d • e

MS I Vs B 3.7.2

/\A I

BASES APPLICABLE SAFETY ANALYSES The design basis of the MSIVs is established by the containment analysis for the Main Steam Line Break (MSLB) 3.i-.z..:-'2-inside containment, as discussed in the FSAR, Section 14.18 _4 (Ref. 2). It is also influenced by the accident analysis of

~---~~-:~~~~~-1fM.....i;i~~~).)}~~~~ep~~~e~~~~ei~i~~~r~~~Rii~~~~~~nc!~~;~ one forrnsLB

Ct.J.:JE.e.T fuel integrity and two for containment analysis (on for "'\[+he..+ hQJX..

containment temperatur~ and one for containment pressure). (b<<n ~~IAffLJ The 1i mit i ng case for t.!ue 1/i ntegr/IHy aiid) containment temperature is the hot fulT power MSLB inside containment following a turbine trip. At hot full power the stored energy in the primary coolant is maximized. 1 e mos reac 1ve con ro ro assume s uc 1n t e fully withdrawn position, there is an increased possibility that the core c.Q.

will return to power. The core is ultimately shut down by combination of doppler feedback, steam generator dryout, an borated water injection delivered b,.l'......;:.:th..:..::e:.-.=.E~m:::..er:...:i.:;e:.:.:n~~1.1..111--

Coo l in S stem.

a nd ~v<l.. 1n~~r1+j The limiting case fo~the containmeHt analysis for containment pressure 1s the hot zero power MSLB inside containment. At zero power, the steam generator inventory and temperature are at their maximum, maximizing the analyzed mass and energy release to the containment.

Reverse flow due to the open MSIV bypass valves, contributes to the total release of the additional mass and energy."----

The accident analysis compares several different MSLB events against different acceptance criteria. The MSLB outside containment upstream of the MSIV is limiting for offsite dose, although a break in this short section of main steam header has a very low probability. The MSLB inside containment at hot full power is the limiting case for a post trip return to power. The analysis includes scenarios with offsite power available and with a loss of offsite power following a turbine trip.

With offsite power available, the primary coolant pumps continue to circulate coolant through the steam generators, maximizing the Primary Coolant System (PCS) cooldown. With a loss of offsite power, the response of mitigating systems, such as the High Pressure Safety Injection (HPSI) pumps, is

.delayed.

- --9904060684- 990330--

PDR ADOCK 05000255 p PDR Palisades Nuclear Plant B 3.7.2-2 01/20/98

INSERT The MSIVs are swing disc check valves. The inherent characteristic of this type of valve allows for reverse flow through the valve on a differential pressure even if the valve is closed.

In the event of an MSLB, if the MSIV associated with the unaffected steam generator fails to close, both steam generators may blowdown. This failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a double steam generator blowdown event, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences.

Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial improvement to plant safety.

SECTION 3.7 INSERT I

..... ass~g the ~ormally closed MSIV bypass valves are closed. The MSIV bypass valves do not receive an Isolation signal and might be open during zero power conditions.

INSERT 2 mSLB / .fna.i hB.v-. b.t..-." Q..\,KJ./l.)O.fd

There are thre~ different lirnitingcase~ one for fuel integrity and two for containment analysis

~ne for contamment te~perature and one for containment pressure). The limiting case for fuel/in~e~fltY a~lcon~~unment temperature is the hot full power MSLB inside containment foll~wI.ng a tur~me tn . At hot full ower, the stored ener y in the rirnary coolant is max~ized._ Ith the most reactive control rod assumed stuck in the fully withdraw 't' there IS an mcre~sed. possibility that the core will return to power. The core is ult~t~~;Is::i* e.d

~~wn.by a c~mbination of doppler feedback, steam generator dry out, and borated water IlJeCtion delivered by the Emergency Core Cooling System. .J, (tttJJ:.J +o INSJlj ( Of) f a{;;e_ f) 3. 1:- ~

The MSIVs are swing disc* check valves. The inherent characteristic of this type of valve allows for reverse flow through the valve on a differential pressure even if the valve is closed. In the event of an MSLB, if the MSIV associated with the unaffected steam generator fails to close, both steam generators may blowdown. This failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a double steam generator blowdown event, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial improvement to plant safety.

B 3.7-7

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.3 Main Feedwater Isolation Valves (MSIVs) [and [MFIV] bypass valves]

3.7.3-1 CTS 4.2, Table 4.2.2, Item 15 ITS 3.7.3, Applicability, Required Action A.I, and Bases Background "'

JFD #1, #4, and #10 Comment: (Contractor comment 3.7.3-1, issue #1) The Bases Background presents the MFRV and MFRV Bypass Valves configuration as a design which is clearly different from the standard design assumed in the STS. The fact that the MFRVs are non-safety and not in safety grade locations should be discussed in the ITS Bases.

Consumers Energv Response:

The Background discussion in the Bases for ITS 3.7.3 has been revised to state that the MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping.

Affected Submittal Pages:

Att 2, ITS 3.7.3, page B 3.7.3-1 Att 5, NUREG 3.7.3, page B 3.7-13 insert 3

MFRVs and MFRV Bypass Valves B 3.7.3

B 3.7 PLANT SYSTEMS B 3.7.3 Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves BASES BACKGROUND The MFRVs and MFRV bypass valves in conjunction with feed pump speed, control Main Feedwater (MFW) flow to the steam generators for level control during normal plant operation.

The valves also isolate MFW flow to the secondary side of the steam generators following a High Energy Line Break (HELB). Closure of the MFRVs and MFRV bypass valves terminates flow to both steam generators. Closure of the MFRV and MFRV bypass valve effectively terminates the addition of feedwater to an affected steam generator, limiting the mass and energy release for Main Steam Line Breaks (MSLBs) inside containment, and reducing the cooldown effects. +o The MFRVs and MFRV bypass valves isolate MFW in(~he event of a secondary side pipe rupture inside containmen&limit the quantity of high energy fluid that enters containment through the break. Controlled addition of Auxiliary Feedwater (AFW) is provided by a separate piping system.

One MFRV and one MFRV bypass valve are located on each MFW line outside containment. The piping volume from the valves to the steam generator must be accounted for in calculating mass and energy releases following an MSLB.

The MFRVs and MFRV bypass valves close on receipt of a isolation signal generated by either; steam generator low pressure from its respective steam generator, or containment high pressure. These isolation signals also actuate the Main Steam Isolation Valves (MSIVs) to close. The MFRVs and MFRV bypass valves may also be actuated manuall . The R s an yp ss va ves a1 as 1 on a oss of air.

However, on y the MFRV~ are equi ed with a handwheel local oper ion. In addition t the MFRVs and MFRV b~

valves, a heck valve outside c *ntainment is availabl to isolate e feedwater line pen trating containment. ~~

A description of the MFRVs and MFRV bypass valves is found 1n the FSAR, Section 10.2.3 (Ref. 1).

rn~e~ Palisades Nuclear Plant B 3.7.3-1 01/20/98

. 3-0...,

INSERT The MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping that fail "as-is" on a loss of air. If required, MFW isolation can be accomplished using manually operated valves located upstream*

or downstream of the MFRVs and MFRV Bypass valves. In addition, each MFRV is equipped with a handwheel that can be used to isolate this MFW flowpath .

.3 -b

SECTION 3.7 INSERT 1

..... and MFRV bypass valves in conjunction with feed pump speed, control Main Feedwater (MFW) flow to the steam generators for level control during normal plant operation. The valves also INSERT 2 Bypass valves fail "as-is" on a ss of air. However, only with a handwheel for local opera on.

The MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping that fail "as-is" on a loss of air. If required, MFW isolation can be accomplished using manually operated valves located upstream or downstream of the MFR Vs and MFRV Bypass valves. In addition, each MFRV is equipped with a handwheel that can be used to isolate this MFW flowpath .

B 3.7-13 3-C...

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

3.7.3-2 SECTION 3.7, PLANT SYSTEMS CTS 4.2, Table 4.2.2, Item 15 ITS 3.7.3, Applicability, Required Action A.1, and Bases

Background

JFD #1, #4, and #10 Comment: (Contractor comment 3.7.3-1, issue #2) The STS is based upon redundant isolation valves in the flow path, in addition to the main feedwater regulating, control or bypass valves on a closed system to containment (three valves in the flow path). The Bases needs additional discussion of the fact that manual valves typically are relied on to isolate the flow paths and that for main feedwater the valves providing containment isolation are check valves.

Consumers Energy Response:

The Background discussion in the Bases for ITS 3.7.3 has been revised to explain that if necessary, main feedwater isolation can be accomplished using manually operated valves lo~ated upstream or downstream of the MFRV~ and MFRV Bypass valves.

The safety analysis described in the ISTS differs from the safety analysis at Palisades. - Specifically, the Background discussion in the Bases for ISTS 3.7.3 discusses the availability of a check valve inside containment used to isolate the feedwater line penetrating containment. In the ISTS, the MFIVs and MFIV Bypass valves are credited in the FWLB analysis. As such, in the event of an FWLB outside containment, the check valves in the feedwater lines provide a leak tight barrier between the steam generators and the ruptured feedwater line outside containment.

In Palisade's safety analysis, an FWLB is not separately analyzed but rather, it is bounded by the Steam Line Rupture Incident, Loss of External Load Event, and Loss of Normal Feedwater Event. Thus, all information in the Bases of ISTS 3.7.3 pertaining to an FWLB has been removed from the Bases of ITS 3.7.3.

The check valves in the feedwater lines near the containment penetration perform a containment isolation function, and as such, are addressed by ITS 3.6.3, "Containment Isolation Valves."

4

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION Affected Submittal Pages:

SECTION 3.7, PLANT SYSTEMS Att 2, ITS 3.7.3, page B 3.7.3-1 Att 2, ITS 3.7.3, page B 3.7.3-2 Att 5, NUREG 3.7.3, page B 3.7-13 insert Att 5, NUREG 3.7.3, page B 3.7-14 5

MFRVs and MFRV Bypass Valves B 3.7.3

B 3.7 PLANT SYSTEMS B 3.7.3 Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves BASES BACKGROUND The MFRVs and MFRV bypass valves in conjunction with feed pump speed, control Main Feedwater (MFW) flow to the steam generators for level control during normal plant operation.

The valves also isolate MFW flow to the secondary side of the steam generators following a High Energy Line Break (HELB). Closure of the MFRVs and MFRV bypass valves terminates flow to both steam generators. Closure of the MFRV and MFRV bypass valve effectively terminates the addition of feedwater to an affected steam generator, limiting the mass and energy release for Main Steam Line Breaks (MSLBs) inside containment, and reducing the cooldown effects'. +o The MFRVs and MFRV bypass valves isolate MFW in(~he event of a secondary side pipe rupture inside containmen~limit the quantity of high energy fluid that enters containment through the brP.ak. Controlled addition of Auxiliary Feedwater (AFW) is provided by a separate piping system .

One MFRV and one MFRV bypass valve are located on each MFW line outside containment. The piping volume from the valves to the steam generator must be accounted for in calculating mass and energy releases following an MSLB.

The MFRVs and MFRV bypass valves close on receipt of a isolation signal generated by either; steam generator low pressure from its respective steam generator, or containment high pressure. These isolation signals also actuate the Main Steam Isolation Valves (MSIVs) to close. The MFRVs MFRV bypass valves may also be actuated manuall . The an --~ffRV"-5yifaTsvaTvesfa l 1 as i ~r-ona-To-ss of a; r.

II However, on)'y the MFRVs are equi ed with a handwheel ~p..\ ,?

local oper 1ion. In addition t the MFRVs and MFRV b ~~ ~

valves, a heck valve outside c ntainment is availabl ~*

isolate e feedwater line pen t_ra_t~cont.:::.a~in.;.;;m.:.:e~n..;..t.:..._,____

A description of the MFRVs and MFRV bypass valves is found in the FSAR, Section 10.2.3 (Ref. 1).

-:r:f\ ~e r-

Palisades Nuclear Plant B 3.7.3-1 01/20/98

INSERT The MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping that fail "as-is" on a loss of air. If required, MFW isolation can be accomplished using manually operated valves located upstream or downstream of the MFRVs and MFRV Bypass valves. In addition, each MFRV is equipped with a handwheel that can be used to isolate this MFW flowpath .

MFRVs and MFRV Bypass Valves B 3.7.3

BASES APPLICABLE SAFETY ANALYSES The MFRVs and MFRV bypass valves satisfy Criterion 3 of 10 CFR 50.36(c)(2).

LCO This LCO ensures that the MFRVs and MFRV bypass valves will isolate MFW flow to the steam generators following an MSLB.

This LCO requires that both MFRVs and both MFRV bypass valves be OPERABLE. The MFRVs and MFRV bypass valves are considered OPERABLE when the isolation times are within limits, and are closed on an isolation signal.

Failure to meet the LCO requirements can result in additional mass and energy being released to containment following an MSLB inside containment.

APPLICABILITY All MFRVs and MFRV bypass valves must be OPERABLE, or either clos~d and deactivated, or isolated by closed manually actuated valves, whenever there is significant mass and energy in the Primary Coolant System and steam generators.

In MODES 1, 2, and 3, the MFRVs or MFRV bypass valves are required to be OPERABLE, except when both MFRVs and both MFRV bypass valves are either closed and deactivated, or isolated by closed manually actuated valves, in order to limit the amount of available fluid that could be added to containment in the case of a secondary system pipe break inside containment. When the valves are either closed and deactivated, or isolated by closed.manually actuated valves, they are already performing their safety function.

Palisades Nuclear Plant B 3.7.3-2 01/20/98

INSERT However, this failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed.

The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a single steam generator blowdown event with continued feedwater, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modification would not provide a cost beneficial improvement to plant safety.

5-d

SECTION 3.7 INSERT 1

..... and MFRV bypass valves in conjunction with feed pump speed, control Main Feedwater (MFW) flow to the steam generators for level control during normal plant operation. The valves also INSERT 2 Bypass valves fail "as-is" on a ss of air. However, only with a handwheel for local opera on.

The MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping that fail "as-is" on a loss of air. If required, MFW isolation can be accomplished using manually operated valves located upstream or downstream of the MFR Vs and MFRV Bypass valves. In addition, each MFRV is equipped with a handwheel that can be used to isolate this MFW flowpath.

B 3.7-13 5- e_

MFIVs [and [MFIV] Byp1ss Vilves)

B 3. 7. ;,

BASES fZJD MFr;N BYP.&..'i'.:. VA1.\/e~

BACKGROUND A description of the HFIVs CiS' found in the FSAR, (continued) Section~lO (Ref. l),_..-----:-----=-=~

z_, '3 p.. AI 3.1. 3 -5 APPLICABLE SAFETY ANALYSES

/O..nau~CJ '" in<. i'7mJl*r-----"l.1 I Con:. rc.VCflJC. Ci)u8) 0.>")C.~' Sir LJ~i

~Al 3.1. 3- Z.

LCO APPLICABILITY The MFIVs and bypass valves must there is significant mass and energy (continued)

CEOG STS B 3.7-14 Rev 1, 04/07/95 C.LOS'E:.'D A~'D 'DE:~1"1\/Ai"!:t:>, OE!. l'SOL.ATi:t>

~ LO$~D Y\Ai-l\JAL.L...Y .Al::.T\JATE:.t:> VA.L~ 1 s- r

SECTION 3.7 INSERT However, this failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a single steam generator blowdown event with continued feedwater, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial improvement to plant safety.

B 3.7-14

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.3-3 CTS 4.2, Table 4.2.2, Item 15 ITS 3.7.3, Applicability, Required Action A.1, and Bases

Background

JFD #1, #4, and #10 Comment: (Contractor comment 3.7~3-1, issue #3) The ITS Applicability is modified to isolate the feedwater line with a "manually actuated valve" which is different from the STS text of "a closed manual valve." The valve difference is not explained and no reason given why this new wording is required. Provide this explanation.

Consumers Energv Response:

"Manually actuated" power operated valves are typically used to isolate the MFRV flow paths rather than "manual valves." These valves are air operated gate valves (CV-0742 and CV-0744) located a few feet down stream of the MFRVs (CV-0701 and CV-0703). They are controlled from the control room. Many manual valves would also be available to isolate the flow path if needed. The air operated gate valves do not isolate the MFRV bypass flow paths. Manual valves would be used if those flow paths needed to be isolated. The existing degree of isolation for the main feedwater lines has previously been found

acceptable by the NRC as transmitted in their Safety Evaluation of February 28, 1986, titled "Main Steam Line Breaks - Single Failures."

Affected Submittal Pag~s:

No page changes.

6

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

3.7.3-4 SECTION 3.7, PLANT SYSTEMS CTS 4.2, Table 4.2.2, Item 15 ITS SR 3.7.3.1 and Bases DOC M.2 and JFD #1 Comment: (Contractor comment 3.7.3-3) ITS SR 3.7.3.1 appears to be acceptable; however, the specific basis for the closure time of 22 seconds should be stated in the SR Bases.

Consumers Energy Response:

The Bases of ITS SR 3.7.3.1 has been revised to clarify that MFRV closure times are bounding values assumed in the MSLB containment response and core response (DNB) analyses. In addition, a reference to the appropriate FSAR Section has been included.

Affected Submittal Pages:

Att 2, ITS B 3.7.3, page B 3.7.3-4 Att 5, NUREG B 3.7.3, page B 3.7-17

7

MFRVs and MFRV Bypass Valves B 3.7.3

BASES SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies the closure time for each MFRV and MFRV bypass valve is ~ 22.0 seconds on an actual or simulated Al actuation signal. Specific signals (e.g., steam generator ~!.,3-4 low pressure and containment high pressure) are tested under b:vncJ,~ Section 3.3. Instrumentation. The MFRV and MFRV bypass 11 11 0 L.QIL>eS valves closure times arel-assumed in the MSLB~containment x

( Re~sj.ci"Ji.J) anal se This SR is normally performed upon returning the x

p ant to operation following a refueling outage. The MFRVs and MFRV bypass valves should not be tested at power since (re.sPonse.. o..nd even a part stroke exercise increases the risk of a valve

( CDrc. fc.S"msc. (OM&) closure with the plant generating power. As these valves are not stroke tested at power, they are exempt from the ASME Code,Section XI (Ref. 2) requirements during operation in MODES 1 and 2.

The Frequency is 18 months. The 18 month *Frequency for valve closure time is based on the refueling cycle.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency .

REFERENCES 1. FSAR, Section 10.2.3

~A\

2. ASME, Boiler and Pressure Vessel Code,Section XI, '3.1.~-11 Inservice Inspection, Article IWV-3400

3. ' f SAf., Jc.CtfOl'J IL/,\~, 'L K

4. f 5~ 1 Se.c:hon 11./ I L/

I x

Palisades Nuclear Plant B 3.7.3-4 01/20/98 1-~

HFIVs [and [HFIV] Bypass Valves]

B 3.7.3 RA* j,7.3-'-/

BAS.ES Cl/o.t."i) /[)

REFERENCES 1. FSAR, Section (c)~(<./li. ,__ V RAI .

~.11 ~ *L(

2. ASME, Boiler and Pressure Vessel Code,Section XI, Inservice Inspection, Article IWV-3400.

~. FSAA I ~<t+/-1M /L/J~46.

,7.

SP&:::1'FI'- ~l(:.,~b.L..S 'E.c... 1 ST1::Afol\ G.~~6.~AT02. Loi...i ~~~u~ At.Jc:>>

CONi-°"1~MEJ.ll ~l<::i~ . P~~\VIZ'E:') A.c& 'T°E."bi-et> IJJJ'bli.ii:.. ~ION

~.~, .I.IJSi~tJ~~A.T.\~IJ. ,,

CEOG STS B 3.7-17 Rev 1, 04/07/95 t-b

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.3-5 ITS Bases Page 3.7.3-2 Comment: The Applicable Safety Analysis states 11 Closure of the MFRVs and

... may also be relied on 11 (emphasis added). Are they or not? That statement does not appear consistent with the LCO discussion which only addresses MSLB.

Consumers Energv Response:

The ITS Applicable Safety Analyses has been revised to clarify that MFRV closure is an initial assumption used in the MSLB analysis, and that valve closure is also relied on to mitigate a steam break for core response analysis. This change establishes consistency with the LCO Bases discussion which states 11

- - valves will isolate MFW flow to the steam generators following an MSLB. 11 Affected Submittal Pages:

Att 2~ ITS B 3.7.3, page B 3.7.3-2 Att 5, NUREG B 3.7.3, page B 3.7-14

8

MFRVs and MFRV Bypass Valves B 3.7.3 BASES APPLICABLE SAFETY ANALYSES following a The MFRVs and MFRV bypass valves satisfy Criterion 3 of '

10 CFR 50.36(c) (2).

LCO This LCO ensures that the MFRVs and MFRV bypass valves will isolate MFW flow to the steam generators following an MSLB.

This LCO requires that both MFRVs and both MFRV bypass valves be OPERABLE. The MFRVs and MFRV bypass valves are considered OPERABLE when the isolation times are within limits, and are closed on an isolation signal.

Failure to meet the LCO requirements can result in additional mass and energy being released to containment following an MSLB inside containment.

APPLICABILITY All MFRVs and MFRV bypass valves must be OPERABLE, or either closed and deactivated, or isolated by closed manually actuated valves, whenever there is significant mass and energy in the Primary Coolant System and steam generators.

In MODES 1, 2, and 3, the MFRVs or MFRV bypass valves are required to be OPERABLE, except when both MFRVs and both MFRV bypass valves are either closed and deactivated, or isolated by closed manually actuated valves, in order to limit the amount of available fluid that could be added to containment in the case of a secondary system pipe break inside containment. When the valves are either closed and deactivated, or isolated by closed manually actuated valves, they are already performing their safety function.

Palisades Nuclear Plant B 3.7.3-2 01/20/98

INSERT

However, this failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed.

The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a single steam generator blowdown event with continued feedwater, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modification would not provide a cost beneficial improvement to plant safety .

8-b

MFIVs (and [MFIV] Byp1ss V1lves]

B 3.7.:,

BACKGROUND (continued)

LCO

~ L.O~E:D YI A~UAl.\..Y At:.TvA:rE:.D VA.L.~ l 8-~

SECTION 3.7

INSERI However, this failure was not analyzed as pare of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a single steam generator blowdown event with continued feedwater, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not pro~ide a cost beneficial improvement to plant safety .

B 3.7-14 8-d

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.3-6 ITS Bases Page 3.7.3-3 Comment: The MFRVs could clearly be closed in less than 8 hrs. The reason 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is the allotted time is to get the plant to the point where conditions support closing the valves. The STS Bases language that was not adopted explained that. That language or something similar would appear to be appropriate.

Consumers Energy Response:

The Bases discussion for ITS 3.7.3 Required Actions A.1 and A.2 has been revised to include language similar to the ISTS that supports the 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time for closing inoperable MFRVs or MFRV Bypass valve.

Affected Submittal Pages:

Att 2, ITS B 3.7.3, page B 3.7.3-3 Att 5, NUREG B 3.7.3, page B 3.7-15

9

MFRVs and MFRV Bypass Valves B 3.7.3 BASES APPLICABILITY Once the valves are closed, deactivation can be accomplished (continued) by the removal of the motive force (e.g., electrical power, air) to the valve to prevent valve opening. Closing another manual valve in the flow path either remotely (i.e., control room hand switch) or locally by manual operation satisfies isolation requirements.

In MODES 4, 5, and 6, steam generator energy is low.

Therefore, the MFRVs and MFRV bypass valves are not required to be OPERABLE.

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each valve.

A.l and A.2 With one MFRV or MFRV bypass valve inoperable, action must be taken to close or isolate the inoperable valve(s) within

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . When these valve(s) are closed or isolated, they are performing their required safety function (e.g., to ~~I .... ~lg

_isolate the line). _

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable to close the MFRV or MFRV bypass valvex> wh1c.h 1rH.lucko f<.rform 1-n a. e..o"'+mli-t.J

~'7 fL:..nt-' Slw+. .clowl\I foo C.sl\Jdi-l-1°"' tho..i /Jup('~J 1sd.a.+1A OT thi. ~t-fc.c.+.:J val"<<!)

B.l and B.2 If the MFRVs or MFRV bypass valves cannot be restored to OPERABLE status, closed, or isolated in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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 .

Palisades Nuclear Plant B 3.7.3-3 01/20/98

HFIVs [and [HFIV] Bypiss Valves]

B 3.7.3 BASES APPLICABILITY (continued)

Q

~ ~F'Fl..v Ai.Jt>

~ V..f:.~V S'l'P~S VA1.v&S AlZ.a.

~l"E.~,

ACTIONS

~h1dl 1ric.lvdc.o pc.~1~

o.. C..On~roil.d f(o,.,, Shut-down

+z:i a. c.c-r-J 1+i&n -+n.s..+

S'vfl (>¢~ 13 ~ i.t.Jm.. o..f -t h<-

Li' f i'..: c. h d (lo/t.t(s)

(continued)

CEOG STS B 3.7-15 Rev 1, 04/07/95

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.4 Atmospheric Dump Valves (ADVs) 3.7.4-1 ITS Bases Page 3.7.4-1 Comment: The Background (as well as the FSAR) state that N2 is not required for operability of the ADVs. What is the technical basis for those statements?

Consumers Energy Response:

The following FSAR excerpt provides the technical basis for the N2 supply to the ADVs:

"There is also a nitrogen backup system installed to provide backup of the instrument air system to the Main Steam System's Atmospheric Dump Valves (ADVs). The nitrogen backup system provides 90 psig ~itrogen to the ADVs provided by the 230 psig bulk nitrogen system. The bulk nitrogen system was utilized as the source of the nitrogen rather than a bottle system due to the large number of bottles that would have been required to provide the four hour coping duration necessary to meet Station Blackout requirements. Use of the bulk nitrogen system allows the ADVs to exceed the Station Blackout coping duration requirements of 10 CFR 50.63 as recommended in Reg Guide 1.155.

See Section 8.1.5 for an additional description of Palisades* response to Station Blackout requirements. The sizing of the pressure regulator provides sufficient nitrogen to fully open all four ADVs simultaneously. There is excess capacity for throttling the ADVs for th~ four hour coping duration.

Nitrogen backup to the ADVs is not considered safety related per Reg Guide 1.155 and is not required for the operability of the ADVs. However, the nitrogen backup is required to support post-fire safe shutdown (Appendix R)."

Furthermore, as discussed in JFD #7 for Specification 3.7.4, the power supplies to the instrument air compressors can be connected to the emergency diesel generators in the event of a loss of offsite power to the station.

This assures a source of motive air will be available to the ADVs in the event of an accident accompanied by a loss of offsite power.

Affected Submittal Pages:

No page changes.

10

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.4-2 ITS Bases Pages 3.7.4.1 &2 JFD #8 Comment: The justification in the JFD for not having a surveillance for the block valve is that no credit is taken for the manual isolation valve in the safety analysis. Following a SGTR, if one ADV is being used to cool the plant and the ADV on the other generator spuriously opens, can the total release be tolerated (assuming the block valve cannot be used to isolate the spuriously open valve)?

Consumers Energv Response:

Yes, the analysis of the radiological consequences for an SGTR event considers the most severe release of secondary as well as primary system activity leaked from the ruptured tube. The SGTR analysis and associated radiological consequences assumes that the faulted S/G is steamed to control level throughout the cooldown. The inventory of fission product activity available for release to the environment is a function of the primary to secondary coolant leakage rate, the assumed fission product concentration, and the mass of the steam discharged to the environment. For an SGTR event concurrent with a loss of offsite power, failure of the ADV on the affected SG to close (or as the result of a spurious reopening) would not result in an increase in the radiological consequences from those previously assumed in the safety.

analysis.

Affected Submittal Pages:

No page changes .

11

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System 3.7.5-1 CTS 3.5.3 ITS 3.7.5 Required Action C.2 DOC L.2 and JFD #1 Comment: (Contractor comment 3.7.5-3, issue #2) The CTS markup indicates that DOC L.2 adds Condition C when CTS 3.5.3 already contains the required Actions C.1 and C.2, as shown on the CTS markup. This apparent contradiction and the CTS markup should be revised to delete the "Adds Condition C11

Consumers Energy Response:

Consumers Energy agrees with the above comment. CTS markup page 3-38a will be revised to delete the notation that Condition C was added.

Affected Submittal Pages:

Att 3, CTS, page 3-38a (ITS 3.7.5, page 2 of 4)

12

c. The fi water makeup o tnt Auxiliar. Feedwater P p Suction (P-8 and P-88) may t inoptrablt f a period of days prov1d d th pump service w er makeup to P. ac. pump P-ac and its responding fl control valve are operable fire COIJb e:, e.

A'P?L.

I 3. 5. 3 RA C.,I

~~!.:Z. §_)

3.5.4 0

Co~t>

~

ti t:>. \

/?Al 3.7.S~/

3*38i Amendment No. 9', 161 August 12, l 994

~E: 2 of= 4

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.5-2 CTS 3.5.4 ITS 3.7.5 Condition D DOC A.8 Comment: (Contractor comment 3.7.5-8) DOC A.8 provides an explanation that

" ... or flow paths ... " are added which does not match the ITS Condition D markup. This apparent inconsistency should be eliminated or further explained in a revised DOC or CTS/ITS markup, as required.

Consumers Energv Response:

Doc A.8 has been revised to eliminate the conflict.

Affected Submittal Pages:

Att 3, DOC 3.7.5, page 2 of 7 13

A.6 SPECIFICATION 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM ATTACHMENT 3 DISCUSSION OF CHANGES A Note has been added to SR 3. 7. 5. 3 which states "Not required to be met in MODES 2 or 3 when AFW is in operation." This Note is needed to prevent unnecessary entering of ACTIONS for LCO 3.7.5 during the startup or shutdown of the plant for not being able to meet the SR. Palisades uses the AFW system for steam generator level control during startup and shutdown in MODES 2, 3, and 4. During these operations the flow control valves used are in manual, and will not open automatically when an actuation signal is received, which would fail the SR. This change is administrative because CTS 4.9.b.l states "each valve to actuates to its correct position (or that the specified flow is established) upon receipt of a simulated auxiliary feedwater pump start signal." During startup or shutdown the valves are providing the proper flow for the existing plant condition. This Note is appropriate because the valves are needed to be throttled in these conditions to prevent overfilling of the steam generators due to low steam flow conditions, also the Note clarifies current licensing basis requirements.

A.7 This change adds the additional "inservice requirements" as described in ASME Code,Section XI to CTS 4.9.a.l and 2. This change is administrative in that these requirements are performed by current surveillances and also this change only combines the two requirements, Code and TS. This change is consistent with NUREG-1432.

A.8 CTS 3.5.4 require!

diate corrective action to restore A in the event that all AFW pumps are inop able. Proposed ITS Condition D ha this same requirement b adds "or flow paths" to the Condition. This change is co idered administrative in at the.-addition clarifi s that the same capability is lost whe there are no AFW flow ths as with no AFW umps. This change maintains consi ncy between NUREG-14 2 .

Palisades Nuclear Plant Page 2of7 01/20/98

INSERT CTS 3.5.4 provides corrective actions in the event all AFW pumps are inoperable. In this case, the capability to provide the required AFW flow to either steam generator has been lost. Propose ITS 3.7.5 Condition D also provides corrective actions when the capability to provide the required AFW flow to either steam generator has been lost. Condition D is stated as 11 two AFW trains inoperable with both steam generators having less than 100% of the AFW flow equivalent to a single Operable AFW train available in Mode 1, 2, or 3 or (the) required AFW train inoperable in Mode 4. 11 Since the AFW system inoperability addressed in ITS 3.7.5 Condition D (a loss of AFW function) is equivalent to the condition presented in CTS 3.5.4, this chang~ has been characterized as administrative in nature .

J 3 -b

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.5-3 CTS 3.5.4 ITS 3.7.5 Condition D, Note to Required Action D.1 JFD #12 Comment: (Contractor comment 3.7.5-10) JFD #12 appears to be acceptable; however, this change should be submitted as a generic STS traveler for approval by the Owners Group and the NRC.

Consumers Energv Response:

Consumers Energy will pose such a TSTF change at the next CE Owners Group Technical Specifications subcommittee meeting.

Affected Submittal Pages:

No page changes .

14

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.5-5 ITS LCD 3.7.5 Comment: The way the CTS is constructed, only one AFW pump could be inoperable at a time meaning that Pump C, which has slightly less performance capability, would never be operable by itself. However, now with the switch to an AFW " train" approach, it is possible that Pump C could be operable by itself which appears to constitute a less restrictive change.

Consumers Energv Response:

Neither the CTS, nor the ITS allow continued operations with only one Operable AFW pump. If two AFW Pumps are inoperable in Modes 1, 2, or 3, proposed ITS Condition C requires the plant to be placed in Mode 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . In Mode 4 only one AFW pump is required to be Operable if the SGs are relied upon for heat removal (this is a new requirement for Palisades). In Mode 4 the maximum average primary coolant temperature is less than 300°F which corresponds to a saturation pressure of 67 psia in the secondary side of the SGs. At this pressure, AFW pump "C" has adequate discharge capacity to fulfill the required heat removal function.

Affected Submittal Pages:

No page changes.

15

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.6 Condensate Storage Tank (CST) 3.7.6-1 ITS Bases 3.7.6 ITS LCO 3.7.6 Comment: The backup water supplies that are required to be verified in the LCO should be discussed in the Bases. FSAR section 9.7.4, #4 states that the primary water storage tank is a backup, when it is in fact required in the TS.

Consumers Energy Response:

Required Action A.1 of ITS 3.7.6 requires a verification that the backup water supplies are Operable whenever the condensate volume is not within limit. The Bases discussion for Required Action A.1 has been revised to clearly indicate that the backup water supply is from the Fire Water System and Service Water System.

The safety analysis assumes that approximately 100,000 gallons of AFW is required to remove decay heat for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following a reactor trip. FSAR Section 9.7.4 item 4 states that the condensate storage tank capacity is 125,000 gallons. However, only about 72,000 gallons of the 125,000 gallons are available to supply the suction of the AFW pumps. Therefore, to ensure an adequate supply of condensate is available, LCO 3.7.6 requires that a combined useable volume of ~ 100,000 gallons be available from both the Condensate Storage Tank and Primary Makeup Storage Tank.

Affected Submittal Pages:

Att 2, ITS B 3.7.6, page B 3.7.6-3 Att 5, NUREG B 3.7.6, page B 3.7-34 16

Condensate Storage and Supply

- - - BASES B 3.7.6

. ACTIONS A.1 and A.2 If the condensate volume is not within the limit, the OPERABILITY of the backup water supplies must be verified by administrative means within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. RI\\

OPERABILITY of the backup feedwater supplies must include 3.1.t.,... \

ve

ication of the OPERABILITY of flow paths from the hrt w~-ttr ~yr-tr.~ ba u s li s to the AFW pumps, and availability of the (

Q.n d S WS water in the backup supplies. The Condensate Storage and Supply volume must be returned to OPERABLE status within 7 days, as the backup supplies may be performing this function in addition to their normal functions. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Com letion Time is reasonable, based on o er tin~

experience, to ver y e o th <&Ji up&afiFl

~~1'9). Additionally, verifying the backup water supplies every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is adequate to ensure the backup water supplies continue to be available. The 7 day Completion Time is reasonable, based on OPERABLE backup water supplies

being available, and the low probability of an event requiring the use of the water from the CST and T-81 occurring during this period.

B.1 and B.2 If the condensate volume cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4, without reliance on steam generator for heat removal, within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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 s1stems

Palisades Nuclear Plant B 3.7.6-3. 01/20/98

!&-a-

CST 8 3.t.6

BASES LCO (continued)

is being relied upon for heat removil,f the is require~

to be OPERABLE. CoAJPe..JS~ :rtl~ IW"O ~I( 'I In NODES 5 and 6, the System is not required.

ACTIONS Fin_ Uh.w S'(s+..,.,.,, a."'J ~ws CEOG STS B 3.7-34 Rev 1, 04/07/95 Ju-6

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REOUESTi SECTION 3.7, PLANT SYSTEMS 3.7.7 Component Cooling Water (CCW) System 3.7.7-1 CTS 3.4.2 and 3.4.3 ITS 3.7.7 Required Action B.2 DOC M.4 Comment: (Contractor comment "3.7.7-3, issue #2) The markup of CTS 3.3.2 appears incorrect because it shows a total of 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months (6+30+24) to place the plant in Cold Shutdown rather than the ITS (and STS) of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Explain this difference. Revise the CTS markup.

Consumers Energv Response:

Consumers Energy agrees, the markup of CTS 3.3.2 has been revised. The words 11 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months have been stricken.

11 Affected Submittal Pages:

Att 3, CTS 3.7.7, page 3-29a (ITS 3.7.7, page 2 of 4) 17

~Y bt modified to inop1r1bl1 for. 1 period o~re injection plJJllO be inoperable ovided operable stat within Z4 hours Ont high-pressu safety tnjectio pump *&y bt inop able provided tht PWIP ts r tortd to operabl status within Z4 ours.

valves, interloc or ptptn9 dir~c y associ1t1d wit one of e above coac>one ind required to unction durin9 1 id1nt condtttons sh1l de111ed to be p t of th1t cet1pon t ind shill .

.. et tht Si81 equirements 1s li 1d for th1t coap ent.

Any valve interlock or pipe s~c~iitfd with t safety inject* n ind sh own coolin9 syst overed under .2e 1bov ut, which is req ed to f1o111ction tn9 1ccid1nt ittons, ~*Y be in 1r1blt for 1 per ~ of no more t n hours .

Amtndmen*t No. ** ;.t, l 72 September 26. 1396 J?

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.7-2 ITS 3.7.7 Bases Comment: It should be in the TS, or at least in the Bases, that the two CCW heat exchangers are both required to be operable to have an operable train of ccw.

Consumers Energv Response:

LCO 3.7.7 requires two Operable trains of CCW. The Bases LCO discussion defines an Operable CCW train as having both CCW heat exchangers Operable.

In addition, the Bases Background discussion also states that both CCW heat exchangers are required for an Operable CCW train.

Affected Submittal Pages:

No page changes .

18

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

3.7.7-3 ITS 3.7.7 Bases SECTION 3.7, PLANT SYSTEMS Comment: With the failure of either valve CV-0945 or 46, one CCW heat exchanger would be inoperable and there would be no operable CCW train as both heat exchangers are required for an operable train. If that is true, how can the Bases make the statement that the system can sustain an active single failure?

Consumers Energy Response:

If either valve CV-0945 or CV-0946 were closed, one CCW heat exchanger would be isolated rendering both trains of CCW inoperable. During the plant conditions when CCW is required to be Operable, both CV-0945 and CV-0946 are in the full open position to supply cooling water to normal plant heat loads.

If a loss of electrical power or control air occurs, both valves fail in the full open position. In the event of a OBA, CV-0945 and CV-0946 receive a confirmatory open signal from the recirculation actuation instrumentation to assure cooling water is available during the recirculation phase of a LOCA.

In this configuration, the function of the CCW System is to provide a gradual reduction in the temperature of the recirculated fluid.

The Bases for the CCW System states, in part, "The CCW System is designed to perform its function with a single failure of any active component, assuming a loss of offsite power." Since CV-0945 and CV-0946 are required to be maintained in the open.position, they do not need to reposition in the event of an accident. As such, an active failure of these valves is not assumed.

Affected Submittal Pages:

No page changes.

19

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.8 Service Water System (SWS)

No Comments.

NRC REQUEST:

3.7.9 Ultimate Heat Sink (UHS)

No Comments.

NRC REQUEST:

3.7 .10 Essential Chilled Water (ECW)

No Comments.

NRC REQUEST:

3.7.11 Control Room Emergency Air Cleanup System (CREACS)

No Comments.

20

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

3.7.12 SECTION 3.7, PLANT SYSTEMS CeRtrel Reem Emer~eRey Air Tem~erature CeRtrel system (CREATES)

Fuel Handling Area Ventilation System 3.7.12-1 CTS 3.8.1 and 3.8.4 ITS 3.7.12 LCO Statement and related Bases No DOC and JFD #4, #5, and #8 Comment: (Contractor comment 3.7.12-1, issue #2) ITS 3.7.12 Bases Background, Insert #2, explains that after the fuel accident occurs, then the operator 11 aligns the fuel handling building exhaust through the emergency filtration arrangement. 11 This appears to be in contradiction with the LCO statement which does not permit any fuel movement until the system is already exhausting through the emergency filtration arrangement. Correct this apparent error or explain the reason for this statement~ Also, explain the operator function to 11 secure 11 a component. Is this to lock in place, open or close these components?

Consumers Energv Response:

The Bases Background discussion* for Specification 3.7.12 has been revised to eliminate the contradictory statement that in the event of a fuel handling or cask drop accident, operators align the fuel handling building exhaust 11 through the emergency filtration arrangement. 11 In addition, the revised wording no longer includes the word secured as it applies to the status or 11 11 condition of plant equipment.

Upon further review of Specification 3.7.12, several inconsistency were identified and corrected. A brief discussion of these changes is provided as follows:

SPECIFICATION LCO The wording of the LCO has been revised to better describe the functional requirement of the specification. That is, the operating fuel handling area exhaust fan must be aligned to the emergency filter bank.

21

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS Consumers Energv Response: (continued)

Actions Table The Actions Table Note has been deleted as discussed in RAI-3.7.12-2. In addition, the Actions have been revised to address the most probable causes for failure to meet the requirements of the LCD. As such, the Actions address the conditions when the Fuel Handling Area Ventilation system is inoperable, not properly aligned, or not in operation. Since the Fuel Handling Area Ventilation system consists of a single train aligned in its accident mitigation mode, the only appropriate Required Actions upon failure to meet the LCD is to immediately suspend all Core Alterations, and suspend movement of fuel assemblies and the fuel cask.

Surveillance Requirements ITS SR 3.7.12.1 has been deleted. The intent of this SR is to ensure the standby FBAC system functions properly. For plants that rely on automatic actuation signals, or whose Applicability includes Modes 1, 2, 3, and 4, performance of this SR fulfill the intended function. However, for the Palisades plant, the Fuel Handling Area Ventilation system is required to be in operation whenever the plant is in the condition specified in the Applicability. Since SRs are only required to be met during the condition specified in the Applicability, performance of a system functional test would be redundant to the requirements of the LCD. Therefore, specifying this SR in the ITS is not necessary.

ITS SR 3.7.12.4 has been deleted. The Fuel Handling Area Ventilation system does not include an automatic actuation feature. The system is manually configured in the emergency filtration mode prior to entering the conditions specified in the Applicability. As such, is not necessary to verify the system bypass damper can be cycled.

BASES

Background

Additional detail has been provided relative to the configuration of the Fuel Handling Building Ventilation system including system operation during normal plant evolutions and during fuel handling and cask movement activities. In addition, the contradictory statement related to system alignment by plant operators has been eliminated (RAI 3.7.12-1).

22

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS Consumers Energy Response: (continued)

Applicable Safety Analyses Additional details have been provided relative to the licensing basis of the fuel handling accident previously approved by the NRC staff, and for the fuel cask drop accident.

LCD Clarifying information relative to the requirements of the LCD has been added.

This includes an explanation of why only one fuel handling exhaust fan is required.

Applicability Clarifying information and an explanation of when the LCD in not applicable (consistent with ISTS format) has been included (RAI 3.7.12-5).

Actions Conforming changes to reflect the revised Specification.

Surveillance Requirements Conforming changes to reflect the revised Specification.

ATTACHMENT 3 DOCs CTS Markup New markups of CTS pages 3-47 and 4-14 have been provided.

Administrative Changes Added new DOC A.4.

More Restrictive Changes Deleted DOCs M.1 and M.3.

Added new DOC M.1.

Renumbered DOC M.4 to M.3.

Less Restrictive Changes-Removal of Details to Licensee Controlled Documents Deleted DOC LA.1.

Renumbered DOC LA.2 to LA.1 with minor wording changes.

Less Restrictive Changes De 1eted DOC L. 1 Added new DOC L.1 (RAI 3.7.12-3) 23

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS Consumers Energy Response: (continued)

ATTACHMENT 4 NSHC Deleted NSHC L.1 Added new NSHC L.1 (RAI 3.7.12-3)

ATTACHMENT 5 NUREG MARKUP New markups provided.*

ATTACHMENT 6 JFDs Deleted JFD 6, and JFDs 8 through 11.

Renumbered JFD 7 to JFD 6 with minor wording changes.

Added new JFDs 7 through 13 (RAI 3.7.12-4 for JFD 9).

Affected Submittal Pages:

In lieu of detailed markups and to facilitate NRC staff review, all pages related to our initial submittal of Specification 3.7.12 have been superseded by this change. The revised pages are provided in Enclosure 3.

24

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.12-2 CTS 3.8.4 ITS 3.7.12 Applicability DOC L.1 and JFD #9 Comment: (Contractor comment 3.7.12-2) The addition of the Action Note is acceptable because the licensee has stated that the Fuel Handling Area Ventilation System does not filter any fission product removal associated with ECCS leakage following an accident. In Modes 1, 2, 3, and 4, the System is independent of reactor operation and is not required to be Operable (that is reflected in the deletion of the Bracketed modes of the ITS markup for Applicability). DOC L.1 justifies the new actions Note based upon how the operator cannot "cease fuel movement" and the need to enter LCO 3.0.3. This appears to be a violation of TS requirements and appears to contradict DOC LA.2. The DOC for this CTS change should be revised. The contents of JFD #9 should be placed in the Bases Applicability discussion, to clearly explain how the Operability of the Fuel Handling Area Ventilation is independent of reactor operations.

Consumers Energv Response:

Consumers Energy agrees~ The proposed Actions Table Note and associated justification have been deleted. The Bases Applicability discussion has been revised to clearly state that the Fuel Handling Area Ventilation system is required whenever the potential exists for an accident which could result in damage to irradiated fuel assemblies.

Affected Submittal Pages:

See RAI 3.7.12-1 25

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.12-3 CTS 4.2 Table 4.2.3, Item 2.c ITS SR 3.7.12.1 DOC LA.1 Comment: (Contractor comment 3.7.12-3) DOC LA.1 states these SR details are moved to the Bases; when in fact, these details are not retained in the ITS SR 3.7.12 Bases. Revise the CTS/ITS markup as applicable to comply with the technical justification as provided in this DOC.

Consumers Energy Response:

Based on the additional restriction of the ITS which requires the Fuel Handling Area Ventilation System to be in the emergency filtration arrangement whenever the plant is in the specified Mode of Applicability, the requirements of CTS 4.2, Table 4.2.3, Item 2.c is no longer necessary. A new DOC (DOC L.1) has been provided to justify this change. As such, the details presently contained in the CTS are no longer required in the Bases for Specification 3.7.12.

Affected Submittal Pages:

See RAI 3.7.12-1 26

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7; PLANT SYSTEMS NRC REQUEST:

3.7.12-4 No CTS requirement STS SR 3.7.14.3 and Bases No JFD Comment: (Contractor comment 3.7.12-5) It is acknowledged that this is not an automatically initiated system; however, the NEI 96-06 guidelines require that all deviations from the STS be justified with a JFD. There is no JFD provided for this STS requirement that is not retained.

Consumers Energy Response:

Consumers Energy agrees. JFD 9 has been provided to describe this deviation from the ISTS.

Affected Submittal Pages:

See RA! 3.7.12-1 27

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.12-5 ITS LCO 3.7.12 Comment: Where does the 90 days of the Applicability come from?

Consumers Energy Response:

The 90 day Applicability is associated with the fuel cask drop analysis. That analysis has shown when irradiated fuel has decayed for 90 days or greater, the dose rates at the site boundary are well within the guideline of 10 CFR 100 for all analyzed cask drop events without crediting filtration by the Fuel Handling Ventilation system. The Bases discussion for the Applicability and Applicable Safety Analyses have been revised to include this information.

Affected Submittal Pages:

See RAI 3.7.12-1 28

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

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.13 ECCS Pump Room Exhaust Air Cleanup System (PREACS) 3.7.13-1 New LCD from CTS Table 3.17.3, Item 4 ITS 3.7~13 LCD Statement and related Bases DOC M.1; JFD #4 and #6 Comment: (Contractor comment 3.7.13-1, issue #2) The ITS markup of the Bases is missing inserts #1, #2, #3, and #4, as identified on page B 3.7-65.

Provide the missing documents.

Consumers Energy Response:

The missing page has been provided as part of this response.

Affected Submittal Pages:

Att 5, NUREG, page B 3.7-65 insert 29

(\J ti C,, ho.,(\ 0w

_ti~c. (..Ui-0 f'O\.IS;nt

. R.Al 3,.7-13*1 SECTION 3.7 INSERT 1

..... isolate the safeguards rooms by closing the inlet and exhaust plenum dampers on the initiation of a high radiation alarm from their respective airborne particulate monitor. This isolation lowers the offsite dose to well within IO CFR 100 (Ref. 1) limits if a leak should occur. Typically, high radiation would only be expected due to excessive leakage during the recirculation phase of operation following a loss of coolant accident (LOCA).

INSERT 2

..... supply plenum damper, an exhaust plenum damper, a radiation monitor, and associated piping, valves, and duct work.

INSERT 3

..... which is addressed in LCO 3.3.10, "Engineered Safeguards Room Ventilation (ESRV)

Instrumentation" INSERT 4

..... shut, isolating the affected safeguards room(s) from the rest of the auxiliary building ventilation system lowering the leakage to the environment from the auxiliary building .

B 3.7-65 o(q-o-

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REOUEST:

3.7.13-2 New LCO from CTS Table 3.17.3, Item 4 ITS 3.7.13 Actions Note DOC M.1 and JFD #2 Comment: (Contractor comment 3.7.13-2) No specific technical justification is provided to explain the rationale for developing this LCO as "Separate Condition entry" rather than as a two train system as the STS is developed.

11 Separate Condition entry is normally used in the STS for individual 11 inoperable components rather than trains. Also, Separate Condition entry" is 11 used where the number of inoperabilities are more than two. Therefore, this does not appear to be an appropriate usage of the Separate Condition entry."

11 The resolution will also depend upon the configuration and contents of each ESRV train noted above in Comment #3.7.13-1.

Consumers Energv Response:

Consumers Energy agrees with the above comment. The Action Note specifying that separate condition entry is allowed for each train has been deleted.

Affected Submittal Pages:

Att l, ITS 3.7.13, page 3.7.13-1 Att 2, ITS B 3.7.13, page B 3.7.13-2 ATT 5, NUREG 3.7.13, page 3.7-29 ATT 5, NUREG B 3.7.13, page B 3.7-67 Att 5, NUREG B 3.7.13, page B 3.7-67 insert 30

ESRV Dampers

3. 7 .13 3.7 PLANT SYSTEMS 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers LCO 3.7.13 Two ESRV Damper trains shall be OPERABLE.

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

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ESRV A.1 Initiate action to Immediately Damper trains isolate ass6ciated inoperable. ESRV Damper train(s).

SURVEILLANCE REQUIREMENTS SURVEILLANCE. FREQUENCY SR 3.7.13.1 Verify each ESRV Damper train closes on an 31 days actual or simulated actuation signal .

Palisades Nuclear Plant 3.7.13-1 Amendment No .. 01/20/98

~0-o-

ESRV Dampers B 3.7.13

BASES LCO Two ESRV Damper trains are required to be OPERABLE to ensure that each engineered safeguards room isolates upon receipt of its respective high radiation alarm. Total system failure could result in the atmospheric release from the engineered safeguards rooms exceeding the required limits in the event of a Design Basis Accident (DBA).

An ESRV Damper train is considered OPERABLE when its associated radiation monitor, instrumentation, ductwork, valves, and dampers are OPERABLE.

APPLICABILITY In MODES 1, 2, 3, and 4, the ESR-Damper trains are required to be OPERABLE consistent with the OPERABILITY requirements of the Emergency Core Cooling System (ECCS).

In MODES 5 and 6, the ESRV Damper trains are not required to be OPERABLE, since the ECCS is not required to be OPERABLE.

A Note has bee a e fo Hie-ACT ONS to clarify the application the Completion me rules. The conditio this Specif cation may be en red independently for ea train. T Completion Time of each inoperable train ill .

be track a separately for* ach train, starting from t e time '

the con ition is entered.

Condition A addresses the failure of one or both ESRV Damper trains. Operation may continue as long as action is .

immediately initiated to isolate the affected engineered safeguards room. With the inlet and exhaust dampers closed, or if the inlet and outlet ventilation plenums are adequately sealed, the engineered safeguards room is isolated and the intended safety function is achieved, since the potential pathway for radioactivity to escape to the environment from the engineered safeguards room has been minimized.

The Completion Time for this Required Action is commensurate with the importance-of m~intaining the engineered safeguards room atmosphere isolated from the outside environment when the ECCS pumps are circulating primary coolant after an accident.

Palisades Nuclear Plant B 3.7.13-2 . 01/~0/98

~ob

£GG$ PBEACSl

3. 7. l 3 3.7 PLANJ-~SYSTEMS 3.7.13 LCO 3.7.13 Two ECCS PREACS trains shall be OPERABLE.

APPLICABIL~TY:

ACTIONS COMPLETION TIME rJ' ("(\Oft.

A. Onf'\ECCS PREACS trainS' inoperable.

A.l B. Req red Action and Be in MOOE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ass ciated Completion Ti not met.

Be in MOOE .

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Oper1t1 each ECCS PRE S train for 31 days

[~ 10 continuous ho s with the heater operating or (for stems without heate

~ 15 minutes].

---~--'---~:.;..::!.!.!.!.:::~.!.:...-'-------.L...--+----~

(continued)

CEOG STS 3.7*29 Rev l, 04/07. 15

-:3D --~

ECCS PREACS B 3.7.13 BASES LCO S 1s consi (continued) compon ts necessa filt ion are OP.

b. f11 ter and strict i ng fl and are ca~ 1t of perf

APPLICABILITY In MOOES l, 2, 3, and 4, the ECCS PREAC required to be OPERABLE cons 1st ant with tht OPERAS I LITY requirements of the EME-f'6Ellk:Y Col!& (ECC~

C.ooc.uJ6 S'VS ~ 1"£AIN5~

In MOOES 5 and 5, the ECCS PREACS not required to be OPERABLE, since tht ECCS is not requ1rtd to be OPERABLE.

ACT IOHS (continued)

CEOG STS B 3.7-67 Rev 1, 04/07/95 30-d

SECTION 3.7 INSERT 1 to the ACTIONS to clarify the applicati of the Completion Time of this Specification may be entered inde ndently for each train. The Completion Times each inoperable train will be tracked se arately for each train, starting from the time the ondition is entered.

INSERT%" I Condition A addresses the failure of one or both ESRV Damper train(s). Operation may continue as long as action is immediately initiated to isolate the affected engineered safeguards room. With the inlet and exhaust dampers closed or if the inlet and outlet ventilation plenums are adequately sealed, the engineered safeguards room is isolated and the intended safety function is achieved, since the potential pathway for radioactivity to escape to the environment from the engineered safeguards room has been minimized.

The Completion Time for this Required Action is commensurate with the importance -of maintaining the engineered safeguards room atmosphere isolated from the outside environment when the ECCS pumps are circulating primary coolant after an accident.

B 3.7-67 30- e..

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.14 Fuel Building Air Cleanup System (FBACS) 3.7.14-1 ITS 3.7.14 Comment: Level is greater than or equal to 674 ft relative to what?

(above MSL)?

Consumers Energy Response:

In general, reference to various plant elevations throughout the CTS, ITS, FSAR, and other plant documents is relative to "mean sea level" and, as such, is not explicitly stated. Since the level of the Great Lakes is currently reported using International Great Lake Datum, discussions pertaining to the .

level of Lake Michigan and to external flooding hazards will specify "mean sea level" as appropriate to clearly indicate the correct reference point (i.e., MSL or !GLD) .

Affected Submittal Pages:

No page changes .

31

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.15 Penetration Room Exhaust Air Cleanup System (PREACS)

No comments NRC REQUEST:

3.7.16 Fuel Storage Pool Water Level 3.7.16-1 New LCO from CTS 5.4.2.c, d, and i; and Table 5.4-1 ITS 3.7.16 LCO statement, SR 3.7.16.1, and Bases JFD #4 Comment: (Contractor comment 3.7.16-1) JFD #4 contains no specific technical justification for not retaining the requirements that spent fuel storage is in accordance with Specification 4.3.1.1. The Bases discussion of LCO and SR 3.7.16.1 state these requirements are met which is in contradiction with the ITS LCO proposed. Provide explanation and technical justification that resolves this apparent inconsistency.

Consumers Energy Response:

A new JFD (JFD #7) has been provided to explain why proposed SR 3.7-.16.1 does not ensure compliance with Specification 4.3.1.1. As such, reference to Specification 4.3.11 in SR 3.7.16.1 can be deleted. Conforming changes have also been made to the Bases to eliminate inconsistency with the actual surveillance requirement.

Affected Submittal Pages:

Att 2, ITS B 3.7.16, page B 3.7.16-2 Att 5, NUREG 3.7.18, page 3.7-39 Att 5, NUREG B 3.7.18, page B 3.7-90 Att 6, JFD 3.7.18, page 1 of 1 32

Spent Fuel Assembly Storage B 3.7.16

BASES.

LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Table 3.7.16-1, in the accompanying LCO, ensures that the kett of the spent fuel pool will always remain < 0.95 assuming the pool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Table 3.7.16-1, in the accompanying LCO. Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Specification 4.3.1.1.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II of"the spent fuel pool ~north tilt pit.

e1~(.f' or -fke.

ACTIONS Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE l, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

A.1 Wheh* the configuration of fuel assemblies stored in Region II the spent fuel pool is not in accordance with Table 3.7.16-1, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Table 3.7.16-1.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS

(?~1'J1.\~ .. \

Palisades Nuclear Plant B 3.7.16-2 01/20/98

Spent Fuel Assembly Storas,!.

a> 3.7.~

3.7 PLANT SYSTEMS 3.7.~ Spent Fuel Assembly Storage the

© ~

. JI..

APPLICABILITY : Whenever any fuel assembly is stored in [Region ()i- ofI\ the s~t fuel storage poo~ v~

o"'fhc. ~ ti I+ fit ACTIONS ('!:'; ....... Al I . -

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A. l l - ---* - --NOTE- -- - -- -- -

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to Immedhtely move the noncomplying

'{Region CZl-f' 1!.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I~

3.7.~.l Verify by administrative means the initial Prior to enrichaient and burnup ~the fuel assembly storing the is in accordance with 3.7.~-l ~ fuel assembly (1) $pee jf i cat iln 4. 3. 111(. . < /{; in ~egion ~t" CD 1t CEOG STS 3.7-39 Rev l, 04/07 . '95

Spent Futl BASES (continued)

APPLICABILITY ACTIONS indic1tin9 thit mov ng rra 1tt ut use w .

LCO 3.0.3 would not specify 1ny 1ct1on. If moving 1rr1dhttd.. fuel userab11es ,.hilt in MOOE 1, 2, 3, or 4, tht*

futl rDOve~nt is independent of re1ctor oper1t1on.

Therefore, in tither c1s1, in1bil1ty to move fuel 1ss1111blits 1s not sufficient r11son to require 1 re1ctor shutdown .

SURYEILLAHCE REQUIRD4EHTS 8)

Cf) i"~lr.

REFERENCES None.

. f r1 If\ to plc;...c,tf\ d

{v ~ j (>.F:h_,,-., hIj I n c.. &t<a1w IT

-QA I ~ 17./"-(

( L~5t:.Rl)

CEOG STS B 3.7-90 Rtv 1, 04/07/95

_ _ _ _ 3d--C-. .

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.7.18,*SPENT FUEL ASSEMBLY STORAGE Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.

1. The brackets have been removed and the proper plant specific information or value has been provided.

2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.

3. The requirement/statement has been deleted since it is not applicable to this facility.

The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5. This change reflects the current licensing basis/technical specification.

6. The storage of failed fuel is accomplished by the use. of canisters that fit in the same storage racks as the fuel assemblies themselves. Therefore, the storage pool does not have any specifically designed rack(s) for failed fuel. The reference to a specific number of storage locations for failed fuel is deleted.

7-.

Palisades Nuclear Plant Page 1of1 01/20/98

INSERT ISTS 3.7.18 applies to plants which restrict the storage of fuel assemblies in high density storage locations based on meeting an acceptable combination of initial enrichment and discharge burnup. For fuel assemblies which do not meet the initial enrichment and discharge burnup requirements, the assemblies may be stored in compliance with other NRC approved methods or configurations as stipulat~d in ISTS 4.3.1.1. ISTS SR 3.7.18.l requires an administrative verification of the initial enrichment and discharge burnup of a fuel assembly prior to storing any assembly in a Region 2 location. For the Palisades Plant, storage of fuel assemblies in high density racks (Region II) is only permitted for fuel assemblies which meet the initial enrichment and discharge burnup requirements. Alternate storage methods or configurations (e.g.,

checkerboading) in Region II has not been approved by the NRC. Therefore, reference to storage of fuel assemblies in accordance with Specification 4.3.1.1 in the LCO, SR, and SR Bases has been deleted.

Assurance that fuel assembly enrichments do not exceed the limits of Region I locations (ITS 4.3.1.1) is controlled administratively in the design of new cores and the procurement of new fuel .

~ 2.-e_

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.16-2 CTS 5.4.2.c and d Bases for ITS 3.7.16 No DOC Comment: (Contractor comment 3.7.16-5) The movement of these CTS requirements to a lo~ation under licensee control must be justified with a DOC as required by NEI 96-06. Provide the necessary technical justification in a "LA" DOC and revise the CTS markup as required.

Consumers Energy Response:

CTS page 5-4a has been provided only to show that a new specification (ITS 3.7.16) has been added. As denoted on this page, the requirements of CTS 5.4.2c and CTS 5.4.2d are addressed in proposed Specification 4.3. The addition of Specification 3.7.16 is justified in DOC M.1.

Affected Submittal Pages:

No page changes.

33

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.16-3 ITS 3~7.16 Applicability Comment: The Applicability would be much clearer if it was written as Region II, of either the SFP or the north tilt pit. The present version could be read as Region II of the SFP or anywhere in the north tilt pit.

Consumers Energy Response:

Consumers Energy agrees with the above comment. The Applicability has been revised as suggested.

Affected Submittal Pages:

Att 1, ITS 3.7.16, page 3.7.16-1 Att 2, ITS B 3.7.16, page B 3.7.16-2 Att 5, NUREG 3.7.18, page 3.7-39 Att 5, NUREG B 3.7.18, page B 3.7-90 34

Spent Fuel Assembly Storage

3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage

3. 7 .16

~A' 3.1. lfo' 3 LCO 3.7.16 The combination of initial enrichment and burnup of each spent fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.

e..1+kr APPLICABILITY: Whenever any fuel assembly is stored in Region II ofJ\the spent fuel pool~ north tilt pit.

or~

ACTIONS

NOTE-----------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME

A. Requirements of the LCO not met.

A. l Initiate action to move the noncomplying fuel assembly from Region I I.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial Prior to enrichment and burnup of the fuel assembly storing the is in accordance with Table 3.7.16-1. fuel assembly in Region II

Palisades Nuclear Plant 3.7.16-1 Amendment No. 01/20/98

~L-j Spent Fuel Assembly Storage 8 3.7.16

BASES LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Table 3.7.16-1, in the accompanying LCO, ensures that the keff of the spent fuel pool will always remain< 0.95 assuming the pool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Table 3.7.16-1, in the accompanying LCO. Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Specification 4.3.1.1.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II of/\ the spent fuel pool ra"'~north

. ~

tilt pit.

e.1 :+'ht r . or -t/.-..e.

ACTIONS Required Action A.l is modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6,

LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE l, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

When the configuration of fuel assemblies stored in Region II the spent fuel pool is not in accordance with Table 3.7.16-1, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Table 3.7.16-1.

SURVEILLANCE SR 3.7.16.1.

REQUIREMENTS

@ii '.) 1.\~" \

Palisades Nuclear Plant B 3.7.16-2 01/20/98 J4-b

Spent Fuel Assembly Stora~~

~ 3.7.~

3.7 Pl.ANT SYSTEMS 3.7.~ Spent Fuel Assembly Storage 0 <ti e'i~<.('

x APPLICABILITY: Whenever any fuel assembly. is stored 1n1Region,of 1 th~ >(

s~1' fuel storage poo~ * .

or +1'2* ~+ii+,,+ AAI 311~/3 x ACTIONS (!:; '-. '

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met.

A. l 1--------NOTE---------

LCO 3.0.3 ls not applicable.

I ---------------------

© Initiate action to move the noncomply;ng fuel assembly from

'{Region~

1t Irmied h te 1y SURVEILLANCE REQUIREMENTS SURVEILLANCE . FREQUENC'f er!

S.l.(:t.J

3. 7 .~.1 Verify by administrative means the initial 1nr1chlllent and burnup ~he fuel assembly i~ in accordance with

3.7.111\-1 ~

Prior to storing the fuel assembly

.5 ecif i cat itn 4. 3. l/U. ~ It, in ,fReg ion ~("'

lt

RA\ 3.1./{s,* J x

CEOG STS 3.7-39 Rev 1, 04/07, 9S

Spent Fuel Assembly Stora~

B 3.7.

BASES (continued) l APPLICABILITY ACTIONS mov ng rra 1t1 ue asse w 1 n .

LCO 3.0.3 would not specify any action. If moving trr1di1ttd fuel 1sselllblies while in HOOE l, Z, 3, or 4, the fuel 110ve111ent ts independent of reactor operation.

Thertfort, in tither case, tnabiltty to move fuel assteblies ts not sufftctent reason to require a reactor shutdown *

SURVEILLANCE REQUIREMENTS 8)

(]) 1"'~1..

REFERENCES None.

pr1 v\ to pla...c,,tl\ d -fh~

.fv~) ~bij 1(/ c... S&Ca I m rr

RA t ~ 17./~-/

( *I.~5t:.Rl')

B 3.7-90 Rev 1, 04/07/95 CEOG STS

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.17 Fuel Storage Pool Boron Concentration 3.7.17-1 CTS 4.2, Table 4.2.1, Item #7 ITS SR 3.7.17.1 DOC L.1 Comment: (Contractor comment 3.7.17-3) The removal of this CTS requirement appears acceptable; however, the DOC L.1 explains this CTS change but does not provide a specific technical justif cation for why this CTS requirement can be deleted. Provide this missing just fication in a revision to the DOC.

Consumers Energv Response:

DOC L.1 has been revised to provide additional justification for the deletion of CTS 4.2, Table 4.2.1, Item #7.

Affected Submittal Pages:

Att 3, DOC 3.7.17, page 2 of 2 Att 4, NSHC 3.7.17, page 1 of 2 35

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY A.5 CTS 3. l.5c requires that with specific activity of the secondary coolant

>0.1 µ.Ci/gram DOSE EQUIVALENT I-131, the plant must be placed in COLD SHUTDOWN. In proposed ITS the term is replaced with MODE 5 (see DOC A.4).

In proposed ITS 3.7.17 Applicability, the Specification is applicable in MODES 1, 2, 3, and 4. Placing the plant in COLD SHUTDOWN in CTS and having the Applicability in MODES 1, 2, 3, and 4 in proposed ITS is basically the same. This change is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M.1 CTS 4.2 Table 4.2.1, item 7a, requires the specific activity of the secondary coolant system to be determined once per 31 days whenever the gross activity determination indicates iodine concentrations greater than 103 of the allowable limit, and once per 6 months whenever the gross activity determination indicates iodine concentrations

below 10 3 of the allowable limit. Proposed ITS SR 3. 7 .17 .1 will require the specific activity to be determined once per 31 days. The proposed ITS SR will not contain the allowance to extend the SR interval to 6 months whenever the gross activity determination indicates iodine concentration below 10% of the allowable limit. This change does not adversely affect safety because the 31 day interval ensures that the specific activity is checked frequently enough to establish a trend to identify secondary activity problems in a timely manner. Deleting an allowance to extend an SR interval constitutes a more restrictive change. This change is consistent with NUREG-1432.

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

There were no "Removal of Details" associated with this specification.

LESS RESTRICTIVE CHANGES (L)

L.1 TS 4.2, Table 4.2.7 re ires a sample of secondary c lant be analyzed for gross radioactivity 3 times ev ry 7 days with a maximum of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between samples. T is RA/ requirement has been eleted. The CTS contains no CO, limiting value, or Requi ed

~.7./l-1 Actions associated w. this requirement in CTS, o y that sampling is required.

change is consider Less Restrictive qecause this ampling requirement is delete .

This change is co istent with NUREG-1432.

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

INSERT CTS 4.2, Table 4.2.1 requires a sample of secondary coolant to be analyzed for gross radioactivity 3 times every 7 days with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples. The CTS contains no LCO, limiting value, or Required Actions for secondary coolant gross radioactivity, only that sampling is required. The intent of this surveillance is to monitor the iodine concentration in the secondary coolant in order to detennine the frequency at which an isotopic analysis for Dose Equivalent I-131 concentration in the secondary coolant is perfonned. The CTS requires an isotopic analysis for Dose equivalent I-131 of the secondary coolant once per 31 days whenever the gross activity indicates iodine concentrations greater than 10% of the allowable limit or, once per 6 months whenever the gross activity detennination indicates iodine concentrations below 10% of the allowable limit. However as discussed in DOC M.1 for this specification, the extended surveillance interval of 6 months for the detennination of Dose Equivalent I-131 in the secondary coolant has been proposed for deletion and that future testing be perfonned every 31 days.

Thus, the need to perfonn sampling of the secondary coolant for gross radioactivity is no longer necessary and has been delete in the ITS~ This change is acceptable since gross radioactivity in the secondary coolant is not evaluated for radiological consequences in any of the accidents assumed in the FSAR, and the concentration of the Dose Equivalent I-131 in the secondary coolant will continue to be detennined at an appropriate frequency. In addition, radiation monitoring instrumentation, controlled in accordance with the Offsite Dose Calculation Manual (e.g., SG blowdown monitors and condenser off gas monitor), is available to monitor increases in the radioactivity levels in the secondary coolant.* This change is consistent with NUREG-1432.

LESS RESTRICTIVE CHANGE SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY

~.1 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CTS 4.2, Table 4.2.7 require a sample of secondary coolant be analyze for gross radioactivity 3 times every 7 ays with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .betwee samples. This RAI requirement has been delete . The CTS contains no LCO, limiting va e, or Required Actio 3.1.17- / associated with this requir ent in CTS, only that sampling is requir Cl. This change is considered Less Restricti e because this sampling requirement is de ted. This change is consistent with NURE 14

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 deletes the sample requirement for gross radioactivity of the secondary coolant. This sample does not have a detrimental impact on the integrity of any plant structure, system, or component. Deletion of this sample requirement will not alter the operation of any plant equipment, or otherwise increase

its failure probability. As such, the probability of occurrence for a previously analyzed accident is not significantly increased.

The consequences of a previously analyzed event are depende~t 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. Gross radioactivity of the secondary coolant is not an initial condition input assumed for any analyzed event.

The amount of Dose Equivalent 1-131 in the secondary coolant is the assumed parameter. The limit requirement for Dose Equivalent 1-131 remains unchanged and the sampling requirement has become more restrictive (see M. l). The deletion of the gross radioactivity sampling requirement does not affect the assumptions of an analyzed event. This change does not affect the performance of any credited equipment since the sample requirement is for an unassumed parameter. 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 1of2 01/20/98 35-c....

INSERT CTS 4.2, Table 4.2.1 requires a sample of secondary coolant to be analyzed for gross radioactivity 3 times every 7 days with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples. The CTS contains no LCO, limiting value, or Required Actions for secondary coolant gross radioactivity, only that sampling is required. The intent of this surveillance is to monitor the iodine concentration in the secondary coolant in. order to determine the frequency at which an isotopic analysis for Dose Equivalent I-131 concentration in the secondary coolant is perfonned. The CTS requires an isotopic analysis for Dose equivalent I-131 of the secondary coolant once per 31 days whenever the gross activity indicates iodine concentrations greater than 10% of the allowable limit or, once per 6 months whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. However as discussed in DOC M.1 for this specification, the extended surveillance interval of 6 months for the detennination of Dose Equivalent 1-131 in the secondary coolant has been proposed for deletion and that future testing be performed every 31 days.

Thus, the need to perform sampling of the secondary coolant for gross radioactivity is no longer necessary and has been delete in the ITS. This change is acceptable since gross radioactivity in the secondary coolant is not evaluated for radiological consequences in any of the accidents assumed in the FSAR, and the concentration of the Dose Equivalent I-131 in the secondary coolant will continue to be determined at an appropriate frequency. In addition, radiation monitoring instrumentation, controlled in accordance with the Offsite Dose Calculation Manual (e.g., SG blowdown monitors and condenser off gas monitor), is available to monitor increases in the radioactivity levels in the secondary coolant. This change is consistent with NUREG-1432.

3 5-d

. ENCLOSURE 2 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION EDITORIAL CHANGES

UHS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCO 3.7.9 The UHS shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 5b~. 2.5 SR 3.7.9.1 Verify water level of UHS is ~ §71.9 ft 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months above mean ~ea level.

SR 3.7.9.2 Verify water temperature of UHS is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

~ 81.5°F.

Palisades Nuclear Plant 3.7.9-1 Amendment No. 01/20/98

AFW System B 3.7.5 B 3.7 PLANT SYSTEMS B 3. 7. 5 . Auxi 1i ary Feedwater' (AFW) System BASES BACKGROUND The AFW System automatically supplies feedwater to the steam generators to remove decay heat fr.om the Primary Coolant System upon the loss of normal feedwater supply. The AFW pumps take suction through a common suction line from the Condensate Storage Tank (CST) (LCO 3. 7.6, "Condensate Storage and Supply") and pump to the steam generator secondary side via two separate and independent flow paths to a common AFW supply header for each steam generator. The steam generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere from the steam generators via the Main Steam Safety Valves (MSSVs) (LCO 3.7.1, "Main Steam Safety Valves (MSSVs)") or Atmospheric Dump Valves (ADVs) (LCO 3.7.4, "Atmospheric Dump Valves (ADVs)"). If the main condenser is available, steam may be released via the turbine bypass valve.

The AFW System consists of two motor driven AFW pumps and one steam turbine driven pump configured into two trains.

One train (A/B) consists of a motor driven pump (P-8A) and the turbine driven pump (P-88) in parallel, the discharges join together to form a common discharge. The A/B train cJ common discharge separates to form two flow paths,(°;'t'hich ~

feed each steam generator via each steam generate~ AFW X penetration. The second motor driven pump (P-BC) feeds both steam generators through separate flow paths via each steam .

generator AFW penetration and forms the other train (C).

The two trains join together at each AFW penetration to form a conman supply to the steam generators. Each AFW pump is capable of providing 100% of the required capacity to the steam generators as assumed in the accident analysis. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system.

Each motor driven AFW pump is powered from an independent Class lE power supply, and feeds both steam generators.

Palisades Nuclear Plant B 3.7.5-1 01/20/98

AFW System B 3.7.5 BASES BACKGROUND The steam turbine driven AFW pump receives steam from either (continued) main steam header upstream of the Main Steam Isolation VaJVe (MSIV). Each of the steam feed lines will supply 100% of the requirements of the turbine driven AFW pump. The steam supply from steam generator E-50A receives anopen signal from the Auxi 1i ary Feedwater Actuation' Si gna 1 (AFAS) instrumentation. The steam supply from steam generator E-508 does not. This steam source is a manual backup. The turbine driven AFW pump feeds both steam generators through the same flow paths as motor driven AFW pump P-8A.

One pump at full flow is sufficient to remove decay heat and cool the plant to Shutdown Cooling (SOC) System entry conditions.

The AFW System supplies feedwater to the steam generators during normal plant startup, shutdown, and hot standby conditions.

The AFW System is designed to supply sufficient water to the steam generator(s) to remove decay heat with steam generator pressure at the setpoint of the MSSVs, with exception of AFW pump P-8C. If AF~Wu p P-8C is used, operator action may be required to eithe rip two of four Primary Coolant Pumps (PCPs), start an ditional AFW pump, or reduce steam generator pressure. This will allow the required flowrates to the steam generators that are assumed in the safety analyses. Subsequently, the AFW System supplies sufficient water to cool the plant to SOC entry conditions, and steam is released through the ADVs, or the turbine bypass valve if the condenser is available.

The AFW System actuates automatically on low steam generator level by an AFAS as described in LCO 3.3.3, "Engineered Safety Feature (ESF) Instrumentation" and 3.3.4, "ESF Logic." The AFAS initiates signals for starting the AFW pumps and repositioning the valves to initiate AFW flow to the steam generators. The actual pump starts are on an"as required" basis. P-8A is started initially, if the pump fails to start, or if the required flow is not established in a specified period of time, P-8C is started. If P-8A and P-8C do not start, or if required flow is not established in a specified period of time, then P-88 is started.

The AFW System is discussed in the FSAR, Section 9.7

(Ref. 1).

Palisades Nuclear Plant 8 3.7.5-2 01/20/98

AFW System B 3.1.5

BASES APPLICABLE The AFW System mitigates the consequences of ~ny event with SAFETY ANALYSES a loss of normal feedwater.

The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual l_!

heat, by delivering at least the minimum required flow rate to the steam generators at pressures corres ondin ~ ~

lowest MSSV set pressure plus 3% wit exception of AFW pump P-8C. If AFW pump P-8C is used, operator action maybe ~ - -

required to either trip two of the four PCPs, start"an additional AFW pump or reduce steam generator pressure. This will allow the required flowrate to the steam generators that are assumed in the safety analyses.

The limiting Design Basis Accident for the AFW System is a loss of normal feedwater.

In addition, the minimum available AFW flow and system characteristics are serious considerationl in the analysis of a small break loss of coolant accident.

The AFW System design is such that it can perform its function following loss of normal feedwate~combined with a loss of offsite power with one AFW pump injecting AFW to one steam generator.

The AFW System satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO This LCO requires that two AFW trains be OPERABLE to ensure that the AFW System will perform the design safety function to mitigate the consequences of accidents that could result in overpressurization of the primary coolant pressure boundary. Three independent AFW pumps, in two diverse trains, ensure availability of residual heat removal capability for all events accompanied by a loss of offsite power and a single failure. This is accomplished by powering two pumps from independent emergency buses. The third AFW pump is powered by a diverse means, a steam driven turbine supplied with steam from a source not isolated by the closure of the MSIVs .

Palisades Nuclear Plant B 3.7.5-3 01/20/98

BASES Condensate Storage and Supply B 3.7.6 APPLICABLE The Condensate Storage and Supply provides condensate to SAFETY ANALYSES remove decay heat and to cool down the plant following all events in the accident analysis, discussed in the FSAR, Chapters 5 and 14. For anticipated operational occurrences and accidents which do not affect the OPERABILITY of the steam generators, the analysis assumption is generally 30 minutes at MODE 3, steaming through the MSSVs followed by a cooldown to Shutdown Cooling (SOC) entry conditions at the design cooldown rate.

The Condensate Storage and Supply satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO To satisfy accident analysis assumptions, the CST and T-81 must contain sufficient cooling water to remove decay heat for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following a reactor trip from 102% RTP. This amount of time allows for cool down of the PCS to SOC entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this the CST and T-81 must retain sufficient water to ensure adequate net positive suction head for the* AFW pumps, and makeup for steaming required to remove decay heat.

The combined CST and T-81 level required is a usable volume of at least 100,000 gallons, which is based on holding the plant in MODE 3 for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , followed by a cooldown to SOC entry conditions at approximately 75°F per hour. This basis CJ u..;~S ~established by the Systematic Evaluation Program. X OPERABILITY of the Condensate Storage and Supply System is determined by maintaining the combined tank levels at or above the minimum required volume.

APPLICABILITY In MODES 1, 2, and 3, and in MODE 4, when steam generator is being relied upon for heat removal, the Condensate Storage and Supply is required to be OPERABLE.

In MODES 5 and 6, the Condensate Storage and Supply is not

required because the AFW System is not required .

Palisades Nuclear Plant B 3.7.6-2 01/20/98

CCW System B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Component Cooling Water (CCW) System BASES BACKGROUND The CCW System provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (OBA) or transient. During normal operation, the CCW System also provides this function for various nonessential components, as well as the spent fuel pool. The CCW System serves as a barrier to the release of radioactive byproducts between potentially radioactive systems and the Service Water System (SWS}, and thus to the environment.

The CCW System consists of three pumps connected in parallel to common suction and discharge headers. The discharge 0.rJ

- X*

header splits into two parallel heat exchangerslthen combines again into a common di stri buti on header RO~ [wA1c~

heat 1oads. A common surge tank provides the necessary net Sv,,licJ positive suction head for the CCW pumps and a surge volume for the system. A train of CCW~~ be that equipment electrically connected to a common safety bus necessary to transfer heat acquired from the various heat loads to the SWS. There are two CCW trains, each associated with a Safeguards Electrical Distribution Train which are described in Specification 3.8.9, "Distribution Systems - Operating. 11 The CCW train associated with the Left Safeguards Electrical Distribution Train consists of two CCW pumps (P-52A, P-52C},

both CCW heat exchangers (E-54A, E-548), the CCW surge tank (T-3), associated piping, valves, and controls for the equipment to perform their safety function. The CCW train associated with the Right Safeguards Electrical Distribution Train consists of one CCW pump (P-528), both CCW heat exchangers (E-54A, E-548), the CCW surge tank (T-3),

associated piping, valves, and controls for the equipment to perfonn their safety function. The pumps*and valves are automatically started upon receipt of a safety injection actuation signal and all essential valves are aligned to their post accident positions. CCW valve repositioning also occurs following a Recirculation Actuation Signal (RAS) which aligns associated valves to provide full cooling to the CCW heat exchangers during the recirculation phase following a design basis Loss of Coolant Accident (LOCA).

Palisades Nuclear Plant 8 3.7.7-1 01/20/98

sws B 3.7.8

BASES SURVEILLANCE SR 3.7.8.1 (continued)

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

SR 3.7.8.2 This SR verifies proper automatic operation of the SWS valves on an actual or simulated actuation signal. Specific signals (e.g., safety injection) are tested under Section 3.3, 11 Instrumentation. 11 If the isolation valve for the noncritical service water header CV-1359) or for containment air cooler VHX-4 is at on (CV-0869) fail to dv(.. t-6 t"'4. diVvSiei\ ~ close, then both trains of SWS are considered inoperable~

. 61 c.~irA. waxv1 .flow ~This Survei 11 ance is not required for valves that are

~ ~ .

locked, sealed, or otherwise secured in the required position under administrative controls. This SR is modified by a Note which states this SR is not required to be met in

. MODE 4. The instrumentation providing the input signal is

not required in MODE 4, therefore, to keep consistency with Section 3.3, 11 Instrumentation, 11 the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

SR J. 7 .8.3 The SR verifies proper automatic operation of the SWS pumps on an actual or simulated actuation signal in the 11 with standby power available 11 mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, "Electrical Power Systems. 11 This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, 11 Instrumentation, 11 the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

Palisades Nuclear Plant B 3.7.8-5 01/20/98

UHS

. B 3. 7 .9 B 3.7 PLANT SYSTEMS B 3.7.9 Ultimate Heat Sink (UHS)

BASES BACKGROUND The UHS provides a heat sink for process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. as well as during normal operation. This is done utilizing the Service Water System (SWS).

The UHS has been defined as Lake Michigan. The two principal functions of the UHS are the dissipation of residual heat after reactor shutdown, and dissipation of residual heat after an accident.

The basic performance requirements are that an adequate Net Positive Suction Head (NPSH) to the SWS pumps be available, and that the design basis temperatures of safety related equipment not be exceeded.

Additional information on the design and operation of the system along with a list of components served can be found in FSAR, Section 9.1 (Ref. 1).

APPLICABLE The UHS is the sink for heat removed from the reactor core SAFETY ANALYSES following all accidents and anticipated operational occurrences in which the plant is cooled down and placed on shutdown cooling. Maximum post accident heat load occurs between 20 to 40 minutes after a design basis Loss of Coolant Accident (LOCA). Near this time, the plant switches from injection to recirculation. and the containment cooling J'G~H systems are required to remove the core decay heat. ~ c~

TAe e~eFatiA~ li~its a~e based on conservative heat transfer eel analyses for the worst case LOCA. FSAR~S ction 14.18 (Ref. 2) and Design Basis Document (DBD 1.02 (Ref. 3) X' provid~the details of the analysis wh1 forms the basis

~tfle operating limits. The assumptions include: worst )t' expected meteorological conditions. conservative uncertainties when calculating decay heat, and the worst case single active failure.

The UHS satisfies Cr-iterion 3 of 10 CFR 50.36(c) (2) .

Palisades Nuclear Plant B 3.7.9-1 01/20/98

The minimum water level of the UHS is based on the NPSH requirements for the SWS pumps.

The NPSH calculation assumes a minimum water level of 4 feet above the bottom of the pump suction bell which corresponds to an elevation of 557.25 ft. Violation of the SWS pump submergence requirement should never become a factor unless the Lake Michigan water level falls below the top of the sluice gate opening which is at elevation 568.25 ft. Early warning of a falling intake water level is provided by the intake structure level alarm. The nominal lake level is approximately 580 ft mean sea level. The minimum water temperature of the UHS is ...

UHS

BASES B 3.7.9 LCD The UHS is required to be OPERABLE. The UHS is considered OPERABLE if it contains a sufficient volume of water at or below the maximum temperature that would allow the SWS to operate without the loss of NPSH, and without exceeding the maximum design temperature of the equipment served by the SWS. To meet this condition, the UHS temperature should not exceed 8l.5°F and the level should not fall below~ ft above mean sea level during normal plant operation.~ SG;i&.2.S APPLICABILITY In MODES 1, 2, 3, and 4, the U~S is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODES 5 and 6, the OPERABILITY requirements of the UHS are determined by the systems it supports .

ACTIONS A.1 and A.2 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 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 3.7.9.1 REQUIREMENTS This SR verifies adequate cooling can be maintained. The level specified also ensures sufficient NPSH is available for operating the SWS pumps. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to the trending of the parameter variations during the applicable MODES. This SR verifies that the UHS water level is ~ §71.& ft above mean x sea level as measur~d within the boundaries(of the intake structure.

S"U.~*.2.5 Palisades Nuclear Plant B 3.7.9-2 01/20/98

SFP Level B 3.7.14 B 3~7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool (SFP) Water Level BASES BACKGROUND The minimum water level in the SFP meets the assumptions of iodine decontamination factors following a fuel handling or cask drop accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the SFP design is given in the FSAR, Section 9.11 (Ref. 1), and the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.4 (Ref. 2). The assumptions of fuel handling and fuel cask drop accidents are given in the FSAR, Section 14.19 and 14.11 (Refs. 3 and 4), respectively .

APPLICABLE SAFETY ANALYSES The minimum water level in the SFP meets the assumptions of fuel handling or fuel cask drop accident analyses described in References 3 and 4 and are consistent with the assumptions of Regulatory Guide 1.25 (Ref. 5). The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to a person at the exclusion area boundary is well within the 10 CFR 100 (Ref. 6) limits.

f"OJ.)oolf\<,O -

A~qrdiRg te Reference ~here is 23 ft of water between the top of the damaged ruel assembly and the fuel pool x surface for a fuel handling or fuel cask drop accident.

This LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single assembly, dropped and lying horizontally on top of the spent fuel racks, there may be < 23 ft of water above the top of the assembly and the surface, by the width of the assembly. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rods fail from a hypothetical maximum drop.

The SFP water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2) .

Palisades Nuclear Plant B 3.7.14-1 01/20/98

1-SFP Boron Concentration B 3.7.15

BASES APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool until a complete spent fuel pool verification of the stored assemblies has been performed following the last movement of fuel assemblies in the spent fuel pool. This LCO does not apply following the verification since the verification would confirm that there are no misloaded fuel assemblies. With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.

ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE l, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown .

A.1. A.2.1. and A.2.2 When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. This does not preclude the movement of fuel assemblies to a safe position. In addition, action must be illlTiedi ately i ni ti ated to restore boron conce*nt~ati to within limit. Alte_!!1ptely, beginning a verificatio f the SFP fuel location~to ensure proper locations of e fuel can be performed.

Palisades Nuclear Plant B 3.7.15-2 01/20/98

Spent Fuel Assembly Storage B 3.7.16

B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage BASES ,

BACKGROUND The spent fuel storage facility is designed to store either new (nonirradiated) nuclear fuel assemblies, or used (irradiated) fuel assemblies in a vertical configuration underwater. The storage pool is sized to store 892 irradiated fuel assemblies, which includes storage for (d failed fuel canisters. The spent fuel storage racks are .

grouped into two regions, Region I and Region II pe~ (S~1Sl'l\1' '(

Figure 3. 7.16-1. The racks are designed as a 6+ft'!'! I 1CA-k~1r'f structure able to withstand seismic events. Region I contains racks in the spent fuel pool having a 10.25 inch center-to-center spacing and a single rack in the north tilt pit having a 11.25 inch by 10.69 inch center-to-center spacing. Region II contains racks in both the spent fuel pool and the north tilt pit having a 9.17 inch center~to-center spacing. Because of the smaller spacing and poison concentration, Region II racks have more limitations for fuel storage than Region I racks. Further information on these limitations can be found in Section 4.0, "Design Features." These limitations (e.g., enrichment, burnup) are sufficient to maintain a keff of ~ 0.95 for spent fuel of ori i

of~ to 4.40%. However, ig er initial nrichment fuel h ssem lies ares ored in the spen fuel pool, the must be [,()

stored in a ch kerboard patter taking into ac unt fuel ~

burnup to mai tain a kett of 0. or less.

APPLICABLE The spent fuel storage facility is designed for SAFETY ANALYSES noncriticality by use of adequate spacing, and "flux trap" construction whereby the fuel assemblies are inserted into neutron absorbing stainless steel cans.

The spent fuel assembly storage satisfies Criterion 2 of 10 CFR 50.36(c)(2).

Palisades Nuclear Plant B 3.7.16-1 01/20/98

Secondary Specific Activity B 3.7.17

B 3.7 PLANT SYSTEMS B 3.7.17 Secondary Specific Activity BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Primary Coolant System (PCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives, and thus is indication of current conditions. During transients, I-131 spikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

This limit is lower than the activity value that might be expected from a 1 gpm tube leak of primary coolant at the limit of 1.0 µCi/gm as assumed in the safety analyses with exception of the control rod ejection analysis which assumes 0.6 gpm. LCD 3.4.13, "PCS Operational LEAKAGE," is more restrictive in that the limit for a primary to secondary tube leak is 0.3 gpm. The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and primary coolant LEAKAGE. Most of the iodine isotopes have short half lives (i.e.,< 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ).

Op~rating a plant at the allowable limits ~o~ld result in a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Exclusion Area Boundary (EAB) exposure well within the 10 CFR 100 (Ref. 1) limits.

Palisades Nuclear Plant B 3.7.17-1 Dl/20/98

Secondary Specific Activity B 3.7.17

BASES APPLICABILITY In MODES 1, 2, 3, and 4, the limits on secondary specific activity apply due to the potential for secondary steam releases to the atmosphere.

In MODES 5 and 6, the steam generators are not being used for heat removal. Both the PCS and steam generators are at low pressure or depressurized, and primary to secondary LEAKAGE is minimal. Therefore, monitoring of secondary specific activity is not required.

ACTIONS A.l and A.2 DOSE EQUIVALENT ~I-131 exceeding the allowable value in the -'I.

(cl secondary coolan 1s an indication of a problem in the PC~

and contributes increased post accident doses. If )(

secondary specific activity cannot be restored to within limits in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in 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 3.7.17.1 REQUIREMENTS This SR ensures that the secondary specific activity is within the limits of the accident analysis. A gamma isotope analysis of the secondary coolant, which determines DOSE EQUIVALENT I-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in primary coolant activity or LEAKAGE. The 31 day Frequency is based on the detection of increasing trends of the level of DOSE EQUIVALENT I-131, and allows for appropriate action to be taken to maintain levels below the LCO limit.

Palisades Nuclear Plant B 3.7.17-3 01/20/98

4.Z

The Control Room Ventilation and Isolation System (and2he fuel~to/a 0 a.J

!Area JE¥*f har;?oa 1/Exhi!Js t/Sysfem I sha 11 be demonstrated to be OPERABLE by the o owl~g tests:

1.

<E.~0

,;. I?

3.1, IO, 'Z.

@ I 2. /At li'ast/§nce/per/ref)?el ifg c#le fix:!

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

~

a.

Sg 3.i.1D.'3 b.

(AtiD 5JC:. 3.7.io.

(ft~D SR~1.I0.3 t>§ Nott) @ L./

4-14 Amendment No. a+.-~.~.

ATTACHMENT3 DISCUSSION OF CHANGES SPECIFICATION 3. 7. 7, COMPONENT COOLING WATER (CCW) SYSTEM ADMINISTRATIVE CHANGES (A)

A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.

Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.

A.2 CTS 3.4.2 and 3.4.3 require that if~o!onent(s) listed in Specification 3.4.1 is inoperable for more than the time sp citied, the plant must be placed in HOT SHUTDOWN. In proposed ITS 3. . Required Action B.l, the CTS term is replaced with MODE 3. This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

cc/._

X A.3 CTS 3.4.4 specifies that valves, interlocks and piping that are directly associated with the "above" (CTS 3.4.1) components shall meet the same requirements as listed for that component. CTS 3 .4.5 specifies that valves, interlocks and piping which is associated with the containment cooling system and not covered by CTS 3 .4.4 may be inoperable for no more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if it is required to function during an accident. These requirements are addressed by the definition of OPERABILITY which requires that all associated equipment be OPERABLE. In the proposed ITS, all equipment in a particular train which'is required to function during an accident must be OPERABLE and all equipment in the train will have the same Completion Time. This is an administrative change since the requirement remains that all equipment in a train of containment cooling must be OPERABLE. This change is consistent ,with NUREG-1432.

Palisades Nuclear Plant . Page 1of5 01/20/98

A.4 SPEC

> ~.4.~, ATTACHMENT 3 DISCUSSION OF CHANGES CATION 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM CTS 3.3.2 and 3.4.3 require that with the Required Action and associated Completion -

X Time not met the plant must be placed in COLD SHUTDOWN. In proposed ITS 3.7.7 Required Action B.2, the CTS term is replaced with MODE 5. This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

A.5 CTS 3.4.3 states " .... Continued power operation with one component out of service shall be as specified in Section 3.4.2, with the permissible period in inoperability starting at the time that the first of the two components became inoperable." This explanatory information on the usage rules of technical specifications is addressed in the proposed ITS Section 1.3, "Completion Times," and does not need to be addressed c.J in the Actions of proposed ITS 3. 7~+nus is considered to be an administrative X-change since the requirements on complying with the completion times is addressed in the proposed ITS. This change is consistent with NUREG-1432.

A. 6 The Note added to proposed SR 3 .7. 7 .1 to aid the operator in the prevention of entering an inappropriate LCO. The Note reminds the operator that loss of CCW flow to a

component may render that component inoperable but does not affect the OPERABILITY of the CCW System. This change is considered administrative that this is a clarifier to the operator to prevent confusion. This change is consistent with NUREG-1432.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M.1 CTS 3.3.1, 3.3.2, 3.4.1, and 3.4.2 establish the Applicability for the various components which comprise the CCW by stating that "the reactor shall not be made critical. ... unless all of the following conditions are met." The Applicability of the CCW in proposed ITS 3.7.7 is MODES 1, 2, 3, and 4. As such, the requirements associated with CTS 3.3.1, 3.3.2, 3.4.1, and 3.4.2 have been revised to be more restrictive by requiring the CCW to also be OPERABLE during the additional MODES 3 and 4. SRs 3.7.7.2 and 3.7.7.3 are modified by a Note which states that these SRs are not required to be met in MODE 4. This is due to the instrumentation providing the signals are not required in MODE 4. This change keeps consistency with ITS 3.3.3, "ESF Instrumentation," and current licensing basis. This change is an additional restriction on plant operations and is consistent with NUREG-1432.

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

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.10, CRV FILTRATION LESS RESTRICTIVE CHANGES (L) 2.

L.1 CTS 4.2, Table 4.2.3, item~a requires a verification that the Control Room Ventilation system automatically switches into the emergency mode of operation on a "containment high pressure and high radiation test signal." The Applicability of this requirement is "above COLD SHUTDOWN, during REFUELING OPERATIONS, during movement of irradiated fuel assemblies, and during movement of a fuel cask in or over the Spent Fuel Pool." Proposed SR 3.7.10.3 requires a verification that each CRV Filtration train actuates on an actual or simulated actuatio~signal. The requirement and Applicability of CTS 4.2, Table 4.2.3, item~ais similar to the requirement and Applicability of SR 3. 7 .10. 3. However, SR 3. 7 .10. 3 is further modified by a Note which states that the SR is "not required to be met during movement of irradiate fuel assemblies in the SFP, or during movement of a fuel cask in or over the SFP. " The purpose of this Note is to exclude the requirement of the SR during those plant evolutions in which no instrumentation is available to actuate the CRV System. The CRV System is designed to automatically switch to the emergency mode of operation on a "containment high pressure or containment high radiation signal." The instruments used to initiate these actuation signals are not capable of

detecting an increase in radiation levels in the fuel handling building, and as such, can not provide automatic actuation of the CRY System in the event of a fuel handling accident or cask drop accident in the SFP. Therefore, the addition of the Note in SR 3. 7 .10. 3 establishes consistency with the design of the CRV System and the requirement of the SR. During movement of irradiate fuel assemblies in the SFP, or during movement of a fuel cask in or over the SFP, manual operator action is necessary to initiate the emergency filtration mode of the CRV System.¢

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

ENCLOSURE 3 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7, PLANT 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 3.7.9-1 ITS 3.7.9-1 03/15/99 Tech change ITS 3.7.12-1 ITS 3.7.12-1 03/15/99 RAI 3.7.12-1 ITS 3. 7 .12-2 ITS 3.7.12-2 03/15/99 RAI 3.7.12-1 ITS 3.7.12-3 ------------ 03/15/99 RAI 3.7.12-1 ITS 3.7.13-1 ITS 3.7.13-1 03/15/99 RAI 3. 7 .13-2 ITS 3.7.16-1 ITS 3.7.16-1 03/15/99 RAI 3.7.16-3 ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL ITS B 3.7.2-2 ITS B 3.7.2-2 03/15/99 RAI 3.7.2-2 ITS B 3 . 7. 2-3 ITS B 3.7.2-3 03/15/99 RAI 3.7.2-2 ITS B 3.7.2-4 ITS B 3.7.2-4 03/15/99 RAI 3.7.2-2 ITS B 3.7.2-5 ITS B 3.7.2-5 03/15/99 RAI 3.7.2-2 ITS B 3.7.3-1 ITS B 3.7.3-1 03/15/99 RAI 3.7.3-1 RAI 3.7.3-2 ITS B 3.7.3-2 ITS B 3.7.3-2 03/15/99 RAI 3.7.3-2 RAI 3.7.3-5 ITS B 3.7.3-3 ITS B 3.7.3-3 03/15/99 RAI 3.7.3-6 ITS B 3. 7. 3-4 ITS B 3.7.3-4 03/15/99 RAI 3.7.3-4 ITS B 3.7.3-5 ITS B 3.7.3-5 03/15/99 RAI 3.7.3-4 ITS B 3. 7. 5-1 ITS B 3.7.5-1 03/15/99 editorial ITS B 3. 7. 5-2 ITS B 3.7.5-2 03/15/99 editorial ITS B 3. 7. 5-3 ITS B 3.7.5-3 03/15/99 editorial ITS B 3.7.6-2 ITS B 3.7.6-2 03/15/99 editorial ITS B 3.7.6-3 ITS B 3.7.6-3 03/15/99 RAI 3.7.6-1 ITS B 3.7 .7-1 ITS B 3.7.7-1 03/15/99 editorial ITS B 3.7.8-5 ITS B 3.7.8-5 03/15/99 . editorial ITS B 3.7.9-1 ITS B 3.7.9-1 03/15/99 Tech change ITS B 3. 7 .9-2 ITS B 3.7.9-2 03/15/99 Tech change ITS B 3.7.9-3 ITS B 3.7.9-3 03/15/99 Tech change ITS B 3.7.12-1 ITS B 3.7 .12-1 03/15/99 RAI 3.7.12-1 ITS B 3. 7 .12-2 ITS B 3.7.12-2 03/15/99 RAI 3.7.12-1 ITS B 3. 7 .12-3 ITS B 3.7.12-3 03/15/99 RAI 3.7.12-1 ITS B 3.7.12-4 ITS B 3.7.12-4 03/15/99 RAI 3.7.12-1 ITS B 3.7.12-5 ITS B 3.7.12-5 03/15/99 RAI 3.7.12-1 ITS B 3.7.12-6 ITS B 3. 7 .12 03/15/99 RAI 3.7.12-1

ITS B 3.7.12-7 03/15/99 RAI 3.7.12-1 1

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7, PLANT SYSTEMS REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT#

ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL (continued)

ITS B 3.7.13-2 ITS B 3.7.13-2 03/15/99 RAI 3. 7 .13-2 ITS B 3.7.14-1 Irs B 3.7.14-1 03/15/99 editorial ITS B 3.7.15-2 ITS B 3.7.15-2 03/15/99 editorial ITS B 3.7.16-1 ITS B 3.7.16-1 03/15/99 editorial ITS B 3.7.16-2 ITS B 3.7.16-2 03/15/99 RA! 3.7.16-1 RA! 3.7.16-3 ITS B 3.7.16-3 03/15/99 RA! 3.7.16-3 ITS B 3.7.17-1 ITS B 3.7.17-1 03/15/99 editorial ITS B 3.7.17-3 ITS B 3.7.17-3 03/15/99 editorial ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL CTS 3.7.5, pg 3-38a CTS 3.7.5, pg 3-38a 03/15/99 RA! 3.7.5-1 CTS 3.7.7, pg 3-29a CTS 3.7.7, pg 3-29a 03/15/99 RA! 3.7.7-1 CTS 3.7.10, pg 4-14 CTS 3.7.10, pg 4-14 03/15/99 editorial CTS 3.7.12, pg 3-47 CTS 3.7.12, pg 3-47 03/15/99 RA! 3.7.12-1 CTS 3.7.12, pg 3-46 CTS 3.7.12, pg 3-46 03/15/99 RA! 3.7.12-1 CTS 3.7.12, pg 4-14 CTS 3.7.12, pg 4-14 03/15/99 RA! 3.7.12-1 DOC 3.7.5, pg 2 of 7 DOC 3.7.5, pg 2 of 7 03/15/99 RAI 3.7.5-2 DOC 3.7.7, pg 1 of 5 DOC 3.7.7, pg 1 of 5 03/15/99 editorial DOC 3.7.7, pg 2 of 5 DOC 3.7.7, pg 2 of 5 03/15/99 editorial DOC 3.7.10, pg 4 of 4 DOC 3.7.10, pg 4 of 4 03/15/99 editorial DOC 3.7.12, pg 1 of 4 DOC 3.7.12, pg 1 of 3 03/15/99 RA! 3.7.12-1 DOC 3.7.12, pg 2 of 4 DOC 3.7.12, pg 2 of 3 03/15/99 RA! 3.7.12-1 DOC 3.7.12, pg 3 of 4 DOC 3.7.12, pg 3 of 3 03/15/99 RA! 3.7.12-1 DOC 3.7.12, pg 4 of 4 --------------------- 03/15/99 RA! 3.7.12-1 DOC 3.7.17, pg 1 of 2 DOC 3.7.17, pg 1 of 3 03/15/99 RA! 3.7.17-1 DOC 3.7.17, pg 2 of 2 DOC 3.7.17, pg 2 of 3 03/15/99 RA! 3.7.17-1

DOC 3.7.17, pg 3 of 3 03/15/99 RA! 3.7.17-1 ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL NSHC 3.7.12, pg 1 of 2 NSHC 3.7.12, pg 1 of 2 03/15/99 RA! 3.7.12-1 NSHC 3.7.12, pg 2 of 2 NSHC 3.7.12, pg 2 of 2 03/15/99 RA! 3.7.12-1 NSHC 3.7.17, pg 1 of 2 NSHC 3.7.17, pg 1 of 2 03/15/99 RA! 3.7.17-1 NSHC 3.7.17, pg 2 of 2 NSHC 3.7.17, pg 2 of 2 03/15/99 RA! 3.7.17-1 ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG 3.7-21 NUREG 3.7-21 03/15/99 Tech change NUREG 3.7-29 NUREG 3.7-29 03/15/99 RA! 3.7.13-2 NUREG 3.7-31 NUREG 3.7-31 03/15/99 RA! 3.7.12-1 NUREG 3.7-31 insert ------------ 03/15/99 RA! 3.7.12-1 NUREG 3.7-31 inserts NUREG 3.7-31 inserts 03/15/99 RA! 3.7.12-1 NUREG 3.7-32 NUREG 3.7-32 03/15/99 RA! 3.7.12-1 NUREG 3.7-33 NUREG 3.7-33 03/15/99 RA! 3.7.12-1 NUREG 3.7-39 NUREG 3.7-39 03/15/99 RA! 3.7.16-1 RA! 3.7.16-3 2

REMOVE PAGES CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7, PLANT SYSTEMS INSERT PAGES REV DATE NRC COMMENT#

ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG B 3.7-7 NUREG B 3.7-7 03/15/99 RAI 3.7.2-2 NUREG B 3.7-7 insert NUREG B 3.7-7 insert 03/15/99 RAI 3. 7.2-2 NUREG B 3.7-8 NUREG B 3.7-8 03/15/99 editorial

NUREG B 3.7-8 insert 03/15/99 editorial NUREG B 3.7-13 insert NUREG B 3.7-13 insert 03/15/99 RAI 3.7.3-1 RAI 3.7.3-2 NUREG B 3.7-14 NUREG B 3.7-14 03/15/99 RAI 3.7.3-2 RAI 3.7.3-5 NUREG B 3.7-14 insert NUREG B 3.7-14 insert 03/15/99 RAI 3.7.3-2 NUREG B 3.7-15 NUREG B 3.7-15 03/15/99 RAI 3.7.3-6 NUREG B 3.7-17 NUREG B 3.7-17 03/15/99 RAI 3.7.3-4 NUREG B 3.7-34 NUREG B 3.7-34 03/15/99 RAI 3.7.6-1 NUREG B 3.7-36 insert NUREG B 3.7-36 insert 03/15/99 RAI 3.7.6-1 NUREG B 3.7-44 insert NUREG B 3.7-44 insert 03/15/99 RAI 3.7.6-1 NUREG B 3.7-47 NUREG B 3.7-47 03/15/99 Tech change

NUREG B 3.7-47 insert 03/15/99 Tech change ,

NUREG B 3.7-49 NUREG B 3.7-49 03/15/99 Tech change

NUREG B 3.7-65 insert 03/15/99 RAI 3.7.13-1 NUREG B 3.7-67 NUREG B 3.7-67 03/15/99 RAI 3.7.13-2 NUREG B 3.7-67 insert NUREG B 3.7-67 insert 03/15/99 RAI 3.7.13-2 NUREG B 3.7-71 NUREG B 3. 7-71 03/15/99 RAI 3.7.12-1 NUREG B 3.7-71 insert NUREG B 3.7-71 insert 03/15/99 RA! 3.7.12-1 NUREG B 3.7-72 NUREG B 3.7-72 03/15/99 RA! 3.7.12-1 NUREG B J.7-72 in. (2pgs) NUREG B 3.7-72 in. (3pgs) 03/15/99 RAI 3.7.12-1 NUREG B 3.7-73 NUREG B 3.7-73 03/15/99 RAI 3.7.12-1 NUREG B 3.7-73 insert NUREG B 3.7-73 insert 03/15/99 RAI 3.7.12-1

. NUREG B 3.7-74 NUREG B 3.7-74 03/15/99 RAI 3.7.12-1 NUREG B 3.7-74 insert -------------- 03/15/99 RAI 3.7.12-1 NUREG B 3.7-75 NUREG B 3.7-75 03/15/99 RAI 3.7.12-1 NUREG B 3.7-75 insert NUREG B 3.7-75 insert 03/15/99 RA! 3.7.12-1 NUREG B 3.7-76 NUREG B 3.7-76 03/15/99 RAI 3.7.12-1 NUREG B 3.7-89 NUREG B 3.7-89 03/15/99 editorial NUREG B 3.7-90 NUREG B 3.7-90 03/15/99 RAI 3.7.16-1 RAI 3.7.16-3 ATTACHMENT 6 TO ITS CONVERSIOH SUBMITTAL JFD 3.7.14, pg 1 of 2 JFD 3.7.14, pg 1 of 4 03/15/99 RAI 3.7.12-1 JFD 3.7.14, pg 2 of 2 JFD 3.7.14, pg 2 of 4 03/15/99 RAI 3.7.12-1

JFD 3.7.14, pg 3 of 4 03/15/99 RAI 3.7.12-1

JFD 3.7.14, pg 4 of 4 03/15/99 RAI 3.7.12-1 JFD 3.7.18, pg 1 of 2 JFD 3.7.18, pg 1 of 2 03/15/99 RAI 3.7.16-1 JFD 3.7.18, pg 2 of 2 JFD 3.7.18, pg ~ of 2 03/15/99 RAI 3.7.16-1 3

UHS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCO 3.7.9 The UHS shall be OPERABLE.

APPLICABILITY: MODES

.. 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify. water level of UHS is ~ 568.25 ft 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months above mean sea level.

SR 3.7.9.2 Verify water temperature of UHS is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

s; 81.5°F .

Palisades Nuclear Plant 3.7.9-1 Amendment No. 03/15/99 J

Fuel Handling Area Ventilation System 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Fuel Handling Area Ventilation System LCO 3.7.12 The Fuel Handling Area Ventilation System shall be OPERABLE with one fuel handling area exhaust fan aligned to the emergency filter bank and in operation.

APPLICABILITY: During movement of irradiated fuel assemblies in the fuel handling building when irradiated fuel assemblies with

< 30 days decay time are in the fuel handling building, During movement of a fuel cask in or over the SFP when irradiated fuel assemblies with < 90 days decay time are in the fuel handling building, During CORE ALTERATIONS when irradiated fuel assemblies with

< 30 days decay time are in the containment with the equipment hatch open, During movement of irradiated fuel assemblies in the containment when irradiated fuel assemblies with

< 30 days decay time are in the containment with the equipment hatch open.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel Handling Area A.1 Suspend movement of Immediately Ventilation System not fuel assemblies.

aligned or in opera ti on. AND OR A.2 Suspend CORE Immediately ALTERATIONS.

Fuel Handling Area Ventilation System AND inoperable.

A.3 Suspend movement of a Immediately fuel cask in or over the SFP.

Palisades Nuclear Plant 3.7.12-1 Amendment No. 03/15/99

Fuel Handling Area Ventilation System 3.7.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Perform required Fuel Handling Area In accordance Ventilation System filter testing in with the accordance with the Ventilation Filter Ventilation Testing Program. Filter Testing Program SR 3.7.12.2 Verify the flow rate of the Fuel Handling 18 months Area Ventilation System, when aligned to the emergency filter bank, is ~ 5840 cfm and :5: 8760 cfm.

Palisades Nuclear Plant 3.7.12-2 Amendment No. 03/15/99

ESRV Dampers 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers LCO 3. 7 .13 Two ESRV Damper trains shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ESRV A.1 Initiate action to Immediately Damper trains isolate associated inoperable. ESRV Damper train(s).

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify each ESRV Damper train closes on an 31 days actual or simulated actuation signal .

Palisades Nuclear Plant 3.7.13-1 Amendment No. 03/15/99

Spent Fuel Assembly Storage 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage LCO 3.7.16 The combination of initial enrichment and burnup of each spent fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.

APPLICABILITY: Whenever any fuel assembly is stored in Region II of*either the spent fuel pool or the north tilt pit.

ACTIONS

NOTE--------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME

A. Requirements of the LCO not met.

A.1 Initiate action to move the noncomplying fuel assembly from Region I I.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial Prior to*

enrichment and burnup of the fuel assembly storing the is in accordance with Table 3.7.16-1. fuel assembly in Region II Palisades Nuclear Plant 3.7.16-1 Amendment No. 03/15/99

MS I Vs B 3.7.2

BASES APPLICABLE The design basis of the MSIVs is established by the SAFETY ANALYSES containment analysis for the Main Steam Line Break (MSLB) inside containment, as discussed in the FSAR, Section 14.18 (Ref. 2). It is also influenced by the accident analysis of the MSLB events presented in the FSAR, Section 14.14 (Ref. 3). The MSIVs are swing disc check valves. The inherent characteristic of this type of valve allows for reverse flow through the valve on a differential pressure even if the valve is closed. In the event of an MSLB, if the MSIV associated with the unaffected steam generator fails to close, both steam generators may blowdown. This failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a double steam generator blowdown event, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences.

Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial improvement to plant safety.

There are three different limiting MSLB cases that have been evaluated, one for fuel integrity 1nd two for containment analysis (one for containment temperature and one for containment pressure). The limiting case for containment temperature is the hot full power MSLB inside containment folfowing a turbine trip. At hot full power, the stored energy in the primary coolant is maximized.

The limiting case for the containment analysis for containment pressure and fuel integrity is the hot zero power MSLB inside containment. At zero power, the steam generator inventory and temperature are at their maximum, maximizing the analyzed mass and energy release to the containment. Reverse flow due to the open MSIV bypass valves, contributes to the total release of the additional mass and energy. With the most reactive control rod assumed stuck in the fully withdrawn position, there is an increased possibility that the core will return to power. The core is ultimately shut down.by a combination of doppler feedback, steam generator dryout, and borated water injection

delivered by the Emergency Core Cooling System.

Palisades Nuclear Plant B 3.7.2-2 03/15/99

MS I Vs B 3.7.2

BASES APPLICABLE The accident analysis compares several different MSLB events SAFETY ANALYSES against different acceptance criteria. The MSLB outside (continued) containment upstream of the MSIV is limiting for offsite dose, although a break in this short section of main steam header has a very low probability. The MSLB inside containment at hot full power is the limiting case for a post trip return to power. The analysis includes scenarios with offsite power available and with a loss of offsite power following a turbine trip.

With offsite power available, the primary coolant pumps continue to circulate coolant through the steam generators, maximizing the Primary Coolant System (PCS) cooldown. With a loss of offsite power, the response of mitigating systems, such as the High Pressure Safety Injection (HPSI) pumps, is delayed.

The MSIVs serve only a safety function and remain open during power operation. These valves operate under the following situations:

a. An MSLB inside containment. For this accident scenario, steam is discharged into containment from both steam generators until closure of the MSIV in the intact steam generator occurs. After MSIV closure, steam is discharged into containment only from the affected steam generator.

b. A break outside of containment and upstream from the MSIVs. This scenario is not a containment pressurization concern. The uncontrolled blowdown of more than one steam generator must be prevented to limit the potential for uncontrolled PCS cooldown and positive reactivity addition. Closure of the MSIVs limits the blowdown to a single steam generator.

c. A break downstream of the MSIVs. This type of break will be isolated by the closure of the MSIVs. Events such as increased steam flow through the turbine or the turbine bypass valve will also terminate on closure of the MSIVs.

d. A steam generator- tube rupture. For this scenario, closure of the MSIVs isolates the affected steam generator from the intact steam generator and minimizes radici1ogical releases .

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

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MS I Vs B 3.7.2

BASES LCO This LCO requires that the MSIV in each of the two steam lines be OPERABLE. The MSIVs are considered OPERABLE when the isolation times are within limits, and they close on an isolation signal.

This LCO provides assurance that the MSIVs will perform their design safety function to mitigate the consequences of accidents that could result in offsite exposures comparable to the 10 CFR 100.11 (Ref. 4) limits or the NRC staff approved licensing basis.

APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3 except when both MSIVs are closed and deactivated when there is significant mass and energy in the PCS and steam generators. When the MSIVs are closed, they are already performing their safety function. Deactivation can be accomplished by the removal of the motive force (e.g., air) to the valve to prevent valve opening.

In MODE 4, the steam generator energy is low; therefore, the MSIVs are not required to be OPERABLE.

In MODES 5 and 6, the steam generators do not contain much energy because their temperature is below the boiling point of water; therefore, the MSIVs are not required for isolation of potential high energy secondary system pipe breaks in these MODES.

ACTIONS With one MSIV inoperable in MODE 1, time is allowed to restore the component to OPERABLE status. Some repairs can be made to the MSIV with the plant hot. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable, considering the probability of an accident occurring during the time period that would require closure of the MSIVs.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is greater than that normally allowed for containment isolation valves because the MSIVs are valves that isolate a closed system penetrating containment.

Palisades Nuclear Plant B 3.7.2-4 03/15/99

MS I Vs B 3.7.2 BASES ACTIONS B.1 (continued)

If the MSIV cannot be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the plant must be placed in~ MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Condition C would be entered. The Completion Time is reasonable, based on operating experience, to reach MODE 2 in an orderly manner and without challenging plant systems.

C.1 and C.2 Condition C is modified by a Note indicating that separate Condition entry is allowed for each MSIV.

Since the MSIVs are required to be OPERABLE in MODES 2 and 3, the inoperable MSIVs may either be restored* to OPERABLE status or closed. When closed, the MSIVs are already in the position required by the assumptions in the safety analysis.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is consistent with that allowed in Condition A.

Inoperable MSIVs that cannot be restored to OPERABLE status within the specified Completion Time, but are closed, must be verified on a periodic basis to be closed. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time js reasonable, based on engineering judgment, MSIV status indications available in the control room, and other administrative controls, to ensure these valves are in the closed position.

D.l and D.2 If the MSIVs cannot be restored to OPERABLE status, or closed, within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply.

To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from MODE 2 in an orderly manner and without challenging plant systems.

Palisades Nuclear Plant B 3.7.2-5 03/15/99

MFRVs and MFRV Bypass Valves B 3.7.3

~ B 3.7 PLANT SYSTEMS B 3.7.3 Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves BASES BACKGROUND The MFRVs and MFRV bypass valves in conjunction with feed pump speed, control Main Feedwater (MFW) flow to the steam generators for level control during normal plant operation.

The valves also isolate MFW flow to the secondary side of the steam generators following a High Energy Line Break (HELB). Closure of the MFRVs and MFRV bypass valves terminates flow to both steam generators. Closure of the MFRV and MFRV bypass valve effectively terminates the addition of feedwater to an affected steam generator, limiting the mass and energy release for Main Steam Line Breaks (MSLBs) inside containment, and reducing the cooldown effects.

The MFRVs and MFRV bypass valves isolate MFW in the event of a secondary side pipe rupture inside containment to limit the quantity of high energy fluid that enters containment through the break. Controlled addition of Auxiliary Feedwater (AFW) is provided by a separate piping system.

One MFRV and one MFRV bypass valve are located on each MFW line outside containment. The piping volume from the valves to the steam ~enerator must be accounted for in calculating-mass and energy releases following an MSLB.

The MFRVs and MFRV bypass valves close on receipt of a isolation signal generated by either; steam generator low pressure from its respective steam generator, or containment high pressure. These isolation signals also actuate the Main Steam Isolation Valves (MSIVs) to close. The MFRVs and MFRV bypass valves may also be actuated manually. The MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping that fail "as-is" on a loss of air. If required, MFW isolation can be accomplished using manually operated valves upstream or downstream of the MFRVs and MFRV Bypass valves.

In addition, each MRFV is equipped with a handwheel that can be used to isolate this MFW fl owpath.

A description of the MFRVs and MFRV bypass valves is found in the FSAR, Section .10.2.3 (Ref. 1) .

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. MFRVs and MFRV Bypass Valves B 3.7.3

BASES APPLICABLE Closure of the MFRVs is an assumption in the MSLB SAFETY ANALYSES containment response analysis. Closure of the MFRVs and MFRV bypass valves is also assumed in the MSLB core response (DNB) analysis.

Failure of an MFRV or MFRV bypass valve to close following a MS~B can result in additional mass and energy to the steam generators contributing to cooldown. This failure also results in additional mass and energy releases following an MSLB event. However, this failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a single steam generator blowdown event with continued feedwater, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not

provide a cost beneficial improvement to plant safety.

The MFRVs and MFRV bypass valves satisfy Criterion 3 of 10 CFR 50.36(c)(2).

LCD This LCD ensures that the MFRVs and MFRV bypass valves will isolate MFW flow to the steam generators following an MSLB.

This LCD requires that both MFRVs and both MFRV bypass valves be OPERABLE. The MFRVs and MFRV bypass valves are considered OPERABLE when the isolation times are within limits, and are closed on an isolation signal.

Failure to meet the LCD requirements can result in additional mass and energy being released to containment following an MSLB inside containment .

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BASES MFRVs and MFRV Bypass Valves B 3.7.3 APPLICABILITY All MFRVs and MFRV bypass valves must be OPERABLE, or either closed and deactivated, or isolated by closed manually actuated valves, whenever there is significant mass and energy in the Primary Coolant System and steam generators.

In MODES 1, 2, and 3, the MFRVs or MFRV bypass valves are required to be OPERABLE, except when both MFRVs and both MFRV bypass valves are either closed and deactivated, or isolated by closed manually actuated valves, in order to limit the amount of available fluid that could be added to containment in the case of a secondary system pipe break inside containment. When the valves are either closed and deactivated, or isolated by closed manually actuated valves, they are already performing their safety function.

Once the valves are closed, deactivation can be accomplished by the removal of the motive force (e.g., electrical power, air) to the valve to prevent valve opening. Closing another

manual valve in the flow path either remotely (i.e., control room hand switch) or locally by manual operation satisfies isolation requirements.

In MODES 4, 5, and 6, steam generator energy is low.

Therefore, the MFRVs and MFRV bypass valves are not required to be OPERABLE.

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each valve.

A.1 and A.2 With one MFRV or MFRV bypass valve inoperable, action must be taken to close or isolate the inoperable valve(s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . When these valve(s) are closed or isolated, they are performing their required safety function (e.g., to isolate the line).

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable to close the MFRV or MFRV bypass valve~ which includes performing a controlled plant shutdown to condition that supports isolation of the affected valve(s).

Palisades Nuclear Plant B 3.7.3-3 03/15/99

MFRVs and MFRV Bypass Valves B 3.7.3

BASES ACTIONS (continued)

B.1 and B.2 If the MFRVs or MFRV bypass valves cannot be restored to OPERABLE status, closed, or isolated in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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 3.7.3.1 REQUIREMENTS This SR verifies the closure time for each MFRV and MFRV bypass valve is ~ 22.0 seconds on an actual or simulated actuation signal. Specific signals (e.g., steam generator low pressure and containment high pressure) are tested under Section 3.3, "Instrumentation." The MFRV and MFRV bypass

valves closure times are bounding values assumed in the MSLB containment response and core response (DNB) analyses (Refs. 3 and 4). This SR is normally performed upon returning the plant to operation following a refueling outage. The MFRVs and MFRV bypass valves should not be tested at power since even a part stroke exercise increases the risk of a valve closure with the plant generating power.

As these valves are not stroke tested at power, they are exempt from the ASME Code,Section XI (Ref. 2) requirements during operation in MODES 1 and 2.

The Frequency is 18 months. The 18 month Frequency for valve closure time is based on the refueling cycle.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency.

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MFRVs and MFRV Bypass Valves B 3.7.3

BASES REFERENCES 1. FSAR, Section 10.2.3

2. ASME, Boiler and Pressure Vessel Code,Section XI, Inservice Inspection, Article IWV-3400

3. FSAR, Section 14.18~2

4. FSAR, Section 14.14

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AFW System B 3.7.5 B 3.7 PLANT SYSTEMS B 3.7.5 Auxiliary Feedwater (AFW) System BASES BACKGROUND The AFW System automatically supplies feedwater to the steam generators to remove decay heat from the Primary Coolant System upon the loss of normal feedwater supply. The AFW pumps take suction through a common suction line from the Condensate Storage Tank (CST) (LCO 3.7.6, 11 Condensate Storage and Supply 11 ) and pump to the steam generator secondary side via two separate and independent flow paths to a common AFW supply header for each steam generator. The steam generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere from the steam generators via the Main Steam Safety Valves (MSSVs) (LCO 3.7.1, 11 Main Steam Safety Valves (MSSVs) 11 ) or Atmospheric Dump Valves (ADVs) (LCO 3.7.4, 11 Atmospheric D.ump Valves (ADVs) 11 ) . If the main condenser is available, steam may be released via the turbine bypass valve.

The AFW System consists of two motor driven AFW pumps and one steam turbine driven pump configured into two trains.

One train (A/B) consists of a motor driven pump (P-8A) and the turbine driven pump (P-8B) in parallel, the discharges join together to form a common discharge. The A/B train common discharge separates to form two flow paths, which feed each steam generator via each steam generator's AFW penetration. The second motor driven pump (P-8C) feeds both steam generators through separate flow paths via each steam generator AFW penetration and forms the other train (C).

The two trains join together at each AFW penetration to form a common supply to the steam generators. Each AFW pump is capable of providing 100% of the required capacity to the steam generators as assumed in the accident analysis. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system.

Each motor driven AFW pump is powered from an independent Class IE power supply, and feeds both steam generators.

Palisades.Nuclear Plant B 3.7.5-1 03/15/99

AFW System B 3.7.5

BASES BACKGROUND The steam turbine driven AFW pump receives steam from either (continued) main steam header upstream of the Main Steam Isolation Valve (MSIV). Each of the steam feed lines will supply 100% of the requirements of the turbine driven AFW pump. The steam supply from steam generator E-50A receives an open signal from the Auxiliary Feedwater Actuation Signal (AFAS) instrumentation. The steam supply from steam generator E-50B does not. This steam source is a manual backup. The turbine driven AFW pump feeds both steam generators through the same flow paths as motor driven AFW pump P-8A.

One pump at full flow is sufficient to remove decay heat and cool the plant to Shutdown Cooling (SOC) System entry conditions.

The AFW System supplies feedwater to the steam generators during normal plant startup, shutdown, and hot standby conditions.

The AFW System is designed to supply sufficient water to the steam generator(s) to remove decay heat with steam generator pressure at the setpoint of the MSSVs, with exception of AFW pump P-8C. If AFW pump P-8C is used, operator action may be required to either trip two of four Primary Coolant Pumps (PCPs), start an additional AFW pump, or reduce steam generator pressure. This will allow the required flowrates to the steam generators that are assumed in the safety analyses. Subsequently, the AFW System supplies sufficient water to cool the plant to SDC entry conditions, and steam is released through the ADVs, or the turbine bypass valve if the *~ondenser is available.

The AFW System actuates automatically on low steam generator level by an AFAS as described in LCO 3.3.3, "Engineered

Safety Feature (ESF) Instrumentation" and 3.3.4, ESF 11 Logic." The AFAS initiates signals for starting the AFW pumps and repositioning the valves to initiate AFW flow to the steam generators. The actual pump starts are on an as 11 required" basis. P-8A is started initially, if the pump fails to start, or if the required flow is not established in a specified period of time, P-8C is started. If P-8A and P-8C do not start, or if required flow is not established in a specified period of time, then P-88 is started .

The AFW System is discussed in the FSAR, Section 9.7 (Ref. 1) .

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AFW System B 3.7.5

BASES APPLICABLE The AFW System mitigates the consequences of any event with SAFETY ANALYSES a loss of normal feedwater.

The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat, by delivering at least the minimum required flow rate to the steam generators at pressures corresponding to the lowest MSSV set pressure plus 3% with the exception of AFW pump P-8C. If AFW pump P-8C is used, operator action may be required to either trip two of the four PCPs, start an additional AFW pump or reduce steam generator pressure. This will allow the required flowrate to the steam generators that are assumed in the safety analyses.

The limiting Design Basis Accident for the AFW System is a loss of normal feedwater.

In addition, the minimum available AFW flow and system characteristics are serious considerations in the analysis of a small break loss of coolant accident *

The AFW System design is such that it can perform its function following loss of normal feedwater combined with a loss of offsite power with one AFW pump injecting AFW to one steam generator.

The AFW System satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO This LCO requires that two AFW trains be OPERABLE to ensure that the AFW System will perform the design safety function to mitigate the consequences of accidents that could result in overpressurization of the primary coolant pressure boundary. Three independent AFW pumps, in two diverse trains, ensure availability of residual heat removal capability for all events accompanied by a loss of offsite power and a single failure. This is accomplished by powering two pumps from independent emergency buses. The

third AFW pump is powered by a diverse means, a steam driven turbine supplied with steam from a source not isolated by the closure of the MSIVs .

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Condensate Storage and Supply B 3.7.6

BASES APPLICABLE The Condensate Storage and Supply provides condensate to SAFETY ANALYSES remove decay heat and to cool down the plant following all events in the accident analysis, discussed in the FSAR, Chapters 5 and 14. For anticipated operational occurrences and accidents which do not affect the OPERABILITY of the steam generators, the analysis assumption is generally 30 minutes at MODE 3, steaming through the MSSVs followed by a cooldown to Shutdown Cooling (SOC) entry conditions at the design cooldown rate.

The Condensate Storage and Supply satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO To satisfy accident analysis assumptions, the CST and T-81 must contain sufficient cooling water to remove decay heat for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following a reactor trip from 102% RTP. This amount of time allows for cool down of the PCS to SOC entry conditions, assuming a coincident lbss of offsite power and the most adverse single failure. In doing this the CST and T-81 must retain sufficient water to ensure adequate net positive suction head for the AFW pumps, and makeup for steaming required to remove decay heat.

The combined CST and T-81 level required is a usable volume of at least 100,000 gallons, which is based on holding the plant in MODE 3 for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , followed by a cooldown to SOC entry conditions at approximately 75°F per hour. This basis was established by the Systematic Evaluation Program. I OPERABILITY of the Condensate Storage and Supply System is determined by maintaining the combined tank levels at or above the minimum required volume.

APPLICABILITY

In MODES 1, 2, and 3, and in MODE 4, when steam generator is being relied upon for heat removal, the Condensate Storage and Supply is required to be OPERABLE.

In MODES 5 and 6, the Condensate Storage and Supply is not required because the AFW System is not required .

Palisades Nuclear Plant B 3;7.6-2 03/15/99

Condensate Storage and Supply B 3.7.6 BASES ACTIONS A.1 and A.2 If the condensate volume is not within the limit, the OPERABILITY of the backup water supplies must be verified by administrative means within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

OPERABILITY of the backup feedwater supplies must include verification of the OPERABILITY of flow paths from the Fire Water System and SWS to the AFW* pumps, and availability of the water in the backup supplies. The Condensate Storage and Supply volume must be returned to OPERABLE status within 7 days, as the backup supplies may be performing this function in addition to their normal functions. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable, based on operating experience, to verify the OPERABILITY of the Fire Water System and SWS. Additionally, verifying the backup water supplies every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is adequate to ensure the backup water supplies continue to be available. The 7 day Completion Time is reasonable, based on OPERABLE backup water supplies being available, and the low probability of an event requiring the use of the water from the CST and T-81 occurring during this period.

B.1 and B.2 If the condensate volume cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4, without reliance on steam generator for heat removal, within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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.

Palisades Nuclear Plant B 3.7.6-3 03/15/99

CCW System B 3.7.7

B 3.7 PLANT SYSTEMS B 3.7.7 Component Cooling Water (CCW) System BASES BACKGROUND The CCW System provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (OBA) or transient. During normal operation, the CCW System also provides this function for various nonessential components, as well as the spent fuel pool. The CCW System serves as a barrier to the release of radioactive byproducts between potentially radioactive systems and the Service Water System (SWS), and thus to the environment.

The CCW System consists of three pumps connected in parallel to common suction and discharge headers. The discharge header splits into two parallel heat exchangers and then combines again into a common distribution header which supplies various heat loads. A common surge tank provides the necessary net positive suction head for the CCW pumps and a surge volume for the system. A train of CCW is considered to be that equipment electrically connected to a common safety bus necessary to transfer heat acquired from the various heat loads to the SWS. There are two CCW trains, each associated with a Safeguards Electrical Distribution Train which are described in Specification 3.8.9, 11 Distribution Systems - Operating. 11 The ccw train associated with the Left Safeguards Electrical Distribution Train consists of two CCW pumps (P-52A, P-52C), both CCW heat exchangers (E-54A, E-548), the CCW surge tank (T-3),

associated piping, valves, and controls for the equipment to perform their safety function. The CCW train associated with the Right Safeguards Electrical Distribution Train consists of one CCW pump (P-528), both CCW heat exchangers (E-54A, E-548), the CCW surge tank (T-3), associated piping, valves, and controls for the equipment to perform their safety function. The pumps and valves are automatically started upon receipt of a safety injection actuation signal and all essential valves are aligned to their post accident positions. CCW valve repositioning also occurs following a Recirculation Actuation Signal (RAS) which aligns associated valves to provide full cooling to the CCW heat exchangers during the recirculation phase following a design basis Loss of Cool ant Accident _(LOCA) .

Palisades Nuclear Plant B 3.7.7-1 03/15/99

sws B 3.7.8

BASES SURVEILLANCE SR 3.7.8.1 (continued)

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

SR 3.7.8.2 This SR verifies proper automatic operation of the SWS valves on an actual or simulated actuation signal. Specific signals (e.g., safety injection) are tested under Section 3.3, "Instrumentation." If the isolation valve for the noncritical service water header (CV-1359) or for containment air cooler VHX-4 (CV-0869) fail to close, then both trains of SWS are considered inoperable due to the diversion of cooling water flow. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

The SR verifies proper automatic operation of the SWS pumps on an actual or simulated actuation signal in the "with standby power available" mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, "Electrical Power Systems." This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Suryeillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint .

Palisades Nuclear Plant B 3.7.8-5 03/15/99

UHS B 3.7.9

B 3.7 PLANT SYSTEMS B 3.7.9 Ultimate Heat Sink (UHS)

BASES BACKGROUND The UHS provides a heat sink for process and operating heat from safety related components during a Design Basis Accident (OBA) or transient, as well as during normal operation. This is done utilizing the Service Water System (SWS).

The UHS has been defined as Lake Michigan. The two principal functions of the UHS are the dissipation of residual heat after reactor shutdown, and dissipation of residual heat after an accident.

The basic performance requirements are that an adequate Net Positive Suction Head (NPSH) to the SWS pumps be available~

and that the design basis temperatures of safety related equipment not be exceeded.

Additional information on the design and operation of the system along with a list of components served can be found in FSAR, Section 9.1 (Ref. 1).

APPLICABLE The UHS is the sink for heat removed from the reactor core SAFETY ANALYSES following all accidents and anticipated operational occurrences in which the plant is cooled down and placed on shutdown cooling. Maximum post accident heat load occurs between 20 to 40 minutes after a design basis Loss of Coolant Accident (LOCA). Near this time, the plant switches from injection to recirculation, and the containment cooling systems are required to remove the core decay heat.

Palisades Nuclear Plant B 3.7.9-1 *a3/15/99

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UHS B 3.7.9 BASES APPLICABLE The minimum water level of the UHS is based on the NPSH SAFETY ANALYSES requirements for the SWS pumps. The NPSH calculation (continued) assumes a minimum water level of 4 feet above the bottom of the pump suction bell which corresponds to an elevation of 557.25 ft. Violation of the SWS pump submergence requirement should never become a factor unless the Lake Michigan water level falls below the top of the sluice gate opening which is at elevation 568.25 ft. Early warning of a falling intake water level is provided by the intake structure level alarm. The nominal lake level is approximately 580 ft mean sea level. The minimum water temperature of the UHS is based on conservative heat transfer analyses for the worst case LOCA. FSAR, Section 14.18 (Ref. 2) and Design Basis Document (DBD) 1.02 (Ref. 3) provide the details of the analysis which forms the basis for these operating limits. The assumptions include: worst expected meteorological conditions, conservative uncertainties when calculating decay heat, and the worst case single active failure.

The UHS satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO The UHS is required to be OPERABLE. The UHS is considered OPERABLE if it contains a sufficient volume of water at or below the maximum temperature that would allow the SWS to operate without the loss of NPSH~ and without exceeding the maximum design temperature of the equipment served by the SWS. To meet this condition, the UHS temperature should not exceed 81.5°F and the level should not fall below 568.25 ft above mean sea level during normal plant operation.

APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODES 5 and 6, the OPERABILITY requirements of the UHS are determined by the systems it supports .

Palisades Nuclear Plant B 3.7.9-2 03/15/99

UHS B 3.7.9 BASES -

ACTIONS A.l and A.2 If the UHS is inoperable, the plant must be placed in a MODE in which the LCD does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 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 3.7.9.1 REQUIREMENTS This SR verifies adequate cooling can be maintained. The level specified also ensures sufficient NPSH is available for operating the SWS pumps. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to the trending of the

parameter variations during the applicable MODES. This SR verifies that the UHS water level is L 568.25 ft above mean sea level as measured within the boundaries of the intake structure.

SR 3.7.9.2 This SR verifies that the SWS is available to provide adequate cooling for the maximum accident or normal -design heat loads following a OBA. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to the trending of the parameter variations during the applicable MODES. This SR verifies that the water temperature from the UHS is

~ 81.5°F.

REFERENCES 1. FSAR, Section 9.1

2. FSAR, Section 14.18

3. Design Basis Document (DBD) 1.02, "Service Water System" Palisades Nucl~ar Plant B 3.7.9-3 03/15/99

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.13-2 New LCO from CTS Table 3.17.3, Item 4 ITS 3.7.13 Actions Note DOC M.1 and JFD #2 Comment: (Contractor comment 3.7.13-2) No specific technical justification is provided to explain the rationale for developing this LCO as 11 Separate Condition entry 11 rather than as a two train system as the STS is developed.

11 Separate Condition entry 11 is normally used in the STS for individual inoperable components rather than trains. Also, 11 Separate Condition entry 11 is used where the number of inoperabilities are more than two. Therefore, this does not appear to be an appropriate usage of the 11 Separate Condition entry. 11 The resolution will also depend upon the configuration and contents of each ESRV train noted above in Comment #3.7.13-1.

Consumers Energv Response:

Consumers Energy agrees with the above comment. The Action Note specifying

that separate condition entry is allowed for each train has been deleted.

Affected Submittal Pages:

Att l, ITS 3.7.13, page 3.7.13-1 Att 2, ITS B 3.7.13, page B 3.7.13-2 ATT 5, NUREG 3.7.13, page 3.7-29 ATT 5, NUREG B 3.7.13, page B 3.7-67 Att 5, NUREG B 3.7.13, page B 3.7-67 insert

30

ESRV Dampers

3. 7 .13

3.7 PLANT SYSTEMS 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers LCO 3. 7.13 Two ESRV Damper trains shall be OPERABLE.

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

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ESRV A.1 Initiate action to Immediately Damper trains isolate associated inoperable . ESRV Damper train(s).

SURVEILLANCE REQUIREMENTS SURVEILLANCE. FREQUENCY SR 3.7.13.1 Verify each ESRV Damper train closes on an 31 days actual or simulated actuation signal.

Palisades Nuclear Plant 3.7.13-1 Amendment No. 01/20/98

~D-.o-

ESRV Dampers B 3.7.13 BASES LCO Two ESRV Damper trains are required to be OPERABLE to ensure that each engineered safeguards room isolates upon receipt of its respective high radiation alarm. Total system failure could result in the atmospheric release from the engineered safeguards rooms exceeding the required limits in the event of a Design Basis Accident (OBA).

An ESRV Damper train is considered OPERABLE when its associated radiation monitor, instrumentation, ductwork, valves, and dampers are OPERABLE.

APPLICABILITY In MODES 1, 2, 3, and 4, the ESR-Damper trains are required to be OPERABLE consistent with the OPERABILITY requirements of the Emergency Core Cooling System (ECCS).

In MODES 5 and 6, the ESRV Damper trains are not required to be OPERABLE, since the ECCS is not required to be OPERABLE.

A Note has bee a e to fne ACT. ONS to c arify the application the Completion me rules. The conditio this Specit cation may be en red independently for ea train. T Completion Time of each inoperable train ill be track a separately for ach train, starting from t e time the con ition is entered.

Conqition A addresses the failure of one or both ESRV Damper trains. Operation may continue as long as action is .

immediately initiated to isolate the affected engineered safeguards room. With the inlet and exhaust dampers closed, or if the inlet and outlet ventilation plenums are adequately sealed, the engineered safeguards room is isolated and the intended safety function is achieved, since the potential pathway for radioactivity to escape to the environment from the engineered safeguards room has been minimized.

The Completion Time for this Required Action is commensurate

. with the importance of maintaining the engineered safeguards room atmosphere isolated from the outside environment when the ECCS pumps are circulating primary coolant after an accident.

Palisades Nuclear Plant B 3.7 .13-2 01/20/98

.30 b

ftG$ PBEACS\

3. 7. 13 3.7 PLAN2-~SYSTEMS 3.7.13 LCO 3.7.13 Two ECCS PREACS trains shall be OPERABLE.

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

- -- N ACTIONS Stf>o.~~

CONDITION COMPLETION TIME t$" rnort-A. One'\ECCS PREACS trainS inoperable.

A. l

8. Req red Action and Be in MOOE 3.

ass ciated Completion Ti not met.

Be in MODE SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Operate each ECCS PRE S train for 31 days

[~ 10 conti~uous ho s with the heater operating or (for stems without heate

~ 15 minutes].

(continued)

CEOG STS 3.7*29 Rev l, 04/07. 15

ECCS PREACS B 3.7.13

BASES LCO (continued) i.

b. filter and stricting fl ind are caP. le of perf

~ APPLICABILITY In MODES l, 2, 3, and 4, the ECCS PREAC required to be OPERABLE consistent with the OPERABILITY requirements of the EME-f'6&k:V C.ol!e c.coc..1~6 S'vST!?M (ECC~ Te>.INt; ~

In MOOES 5 and 6, the ECCS PREACS not required to be OPERABLE, since the ECCS is not required to be OPERABLE *

ACTIONS (continued)

CEOQ STS B 3.7-67 Rev 1, 04/07/95

30-d

SECTION 3.7 INSERT 1 to the ACTIONS to clarify the applicati of the Completion Time of this Specification may be entered inde ndently for each train. The Completion Times each inoperable train will be tracked se arately for each train, staning from the time the ondition is entered.

INSERTr I Condition A addresses the failure of one or both ESRV Damper train(s). Operation may continue as long as action is immediately initiated to isolate the affected engineered safeguards room. With the inlet and exhaust dampers closed or if the inlet and outlet ventilation plenurns are adequately sealed, the engineered safeguards room is isolated and the intended safety function is achieved, since the potential pathway for radioactivity to escape to the envirorunent from the engineered safeguards room has been minimized.

The Completion Time for this Required Action is commensurate with the importance of maintaining the engineered safeguards room atmosphere isolated from the outside envirorunent when the ECCS pumps are circulating primary coolant after an accident .

B 3.7-67 30- e...

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.14 Fuel Building Air Cleanup System (FBACS) 3.7.14-1 ITS 3.7.14 Comment: Level is greater than or equal to 674 ft relative to what?

(above MSL)?

Consumers Energv Response:

In general, reference to various plant elevations throughout the CTS, ITS, FSAR, and other plant documents is relative to "mean sea level" and, as such, is not explicitly stated. Since the level of the Great Lakes is currently reported using International Great Lake Datum, discussions pertaining to the.

level of Lake Michigan and to external flooding hazards will specify "mean sea level" as appropriate to clearly indicate the correct reference point (i.e., MSL or IGLD).

Affected Submittal Pages:

No page changes .

31

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.15 Penetration Room Exhaust Air Cleanup System (PREACS)

No comments NRC REQUEST:

3.7.16 Fuel Storage Pool Water Level 3.7.16-1 New LCO from CTS 5.4~2.c, d, and i; and Table 5.4-1 ITS 3.7.16 LCO statement, SR 3.7.16.1, and Bases JFD #4 Comment: (Contractor comment 3.7.16-1) JFD #4 contains no specific technical justification for not retaining the requirements that spent fuel storage is in accordance with Specification 4.3.1.1. _The Bases discussion of LCO and SR 3.7.16.1 state these requirements are met which is in contradiction with the ITS LCO proposed. Provide explanation and technical justification that resolves this apparent inconsistency .

Consumers Energv Response:

A new JFD (JFD #7) has been provided to explain why proposed SR 3.7.16.l does not ensure compliance with Specification 4.3.1.1. As such, reference to Specification 4.3.11 in SR 3.7.16.1 can be deleted. Conforming changes have also been made to the Bases to eliminate inconsistency with the actual surveillance requirement.

Affected Submittal Pages:

Att 2, ITS B 3.7.16, page B 3.7.16-2 Att 5, NUREG 3.7.18, page 3.7-39 Att 5, NUREG B 3.7.18, page B 3.7-90 Att 6, JFD 3.7.18, page 1 of 1 32

Spent Fuel Assembly Storage B 3.7.16 BASES LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Table 3.7.16-1, in the accompanying LCO, ensures that the kett of the spent fuel pool will always remain < 0.95 assuming the pool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Table 3.7.16-1, in the accompanying LCO. Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Specification 4.3.1.1.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II of/\the spent fuel pool ~north tilt pit.

e.t~(.r

or -the.

ACTIONS Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

When* the configuration of fuel assemblies stored in Region II the spent fuel pool is not in accordance with Table 3.7.16-1, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Table 3.7.16-1.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS

((f\1'J1.\~ .. \

Palisades Nuclear Plant B 3.7.16-2 01/20/98

Spent Fuel Assembly Storai~

. Q) 3.7.'tt

3.7 PLANT SYSTEMS 3.7.~ Spent Fuel Assembly Storage the

APPLICAB IL ITV: Whenever any fuel assembly is stored in [Region~ of!\ the sPc.nt fuel storage poo' .

o<'-tnc. ~ +1 I+ fit vfk<<.

ACTIONS ('f:; ....... RAl :3 1 .lfu -3 I

CONDITION

- " ~

REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met.

A. l """ I - - * - - - - -NOTE- -- - - -- --

LCO 3.0.3 is not applicable.

Initiate action to Inmedhtely move the noncomplying

'{Region~

tr.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I~

crS 3.7.~.l Verify by administrative means the initial Prior to enrichment and burnup ~the fuel assembly storing the s.~.t.J is in accordance with

3. 7 .00-1 ~ fuel assemblt,v Q) 5pec if i cat ifn 4. 3. l/lf. <. 'rt, in .(Region ei1 CD 'lt CEOG STS 3.7-39 Rev 1, 04/07:'95 3z'-b

Spent Fuel BASES (cont;nued)

APPLICABILITY ACTIONS indiciting thit mov ng rra ate ue asse w e .

LCO 3.0.3 would not specify any ict1on. If moving irradiittd. fuel isselllblies while jn MOOE 1, 2, 3, or 4, the fuel rDOve~nt is independent of reictor operation.

Therefore, in either Cise, inability to move fuel asselll011es is not sufficient reison to require i reictor shutdown *

SURYEILLAHCE REQUIREMENTS rJ) i"AA. le.

REFERENCES None.

f rt If. to plu.. Ctl (\ d

_fve.~ ~-"" b/j I() C.. &s<a1ro rr

RA t ~ 17./fo~ I

( L~5t:Rl')

CEOG STS B 3.7-90 Rev 1, 04/07/95 3~-c..

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.7.18, SPENT FUEL ASSEMBLY STORAGE Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS. " The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.

1. The brackets have been removed and the proper plant specific information or value has been provided.

2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.

3. The requirement/statement has been deleted since it is not applicable to this facility.

The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5. This change reflects the current licensing basis/technical specification.

6. The storage of failed fuel is accomplished by the use. of canisters that fit in the same storage racks as the fuel assemblies themselves. Therefore, the storage pool does not have any specifically designed rack(s) for failed fuel. The reference to a specific number of storage locations for failed fuel is deleted.

7-.

Palisades Nuclear Plant Page 1of1 01120/98 3~-d

INSERT ISTS 3.7.18 applies to plants which restrict the storage of fuel assemblies in high density storage locations based on meeting an acceptable combination of initial enrichment and discharge burnup. For fuel assemblies which do not meet the initial enrichment and discharge burnup requirements, the assemblies may be stored in compliance with other NRC approved methods or configurations as stipulated in ISTS 4.3.1.1. ISTS SR 3.7.18.1 requires an administrative verification of the initial enrichment and discharge burnup of a fuel assembly prior to storing any assembly in a Region 2 location. For the Palisades Plant, storage of fuel assemblies in high density racks (Region II) is only permitted for fuel assemblies which meet the initial enrichment and discharge burnup requirements. Alternate storage methods or configurations (e.g.,

checkerboading) in Region II has not been approved by the NRC. Therefore, reference to storage of fuel assemblies in accordance with Specification 4.3.1.1 in the LCO, SR, and SR Bases has been deleted.

Assurance that fuel assembly enrichments do not exceed the limits of Region I locations (ITS 4.3.1.1) is controlled administratively in the design of new cores and the procurement of new fuel .

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.16-2 CTS 5.4.2.c and d Bases for ITS 3.7.16 No DOC Comment: (Contractor comment 3.7.16-5) The movement of these CTS requirements to a location under licensee control must be justified with a DOC as required by NEI 96-06. Provide the necessary technical justification in a 11 LA 11 DOC and revise the CTS markup as required.

Consumers Energy Response:

CTS page 5-4a has been provided only to show that a new specification (ITS 3.7.16) has been added. As denoted on this page, the requirements of CTS 5.4.2c and CTS 5.4.2d are addressed in proposed Specification 4.3. The addition of Specification 3.7.16 is justified in DOC M.1.

Affected Submittal Pages:

No page changes.

33

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION NRC REQUEST:

SECTION 3.7, PLANT SYSTEMS 3.7.16-3 ITS 3.7.16 Applicability Comment: The Applicability would be much clearer if it was written as Region II, of either the SFP or the north tilt pit. The present version could be read as Region II of the SFP or anywhere in the north tilt pit.

Consumers Energy Response:

Consumers Energy agrees with the above comment. The Applicability has been revised as suggested.

Affected Submittal Pages:

Att l, ITS 3.7.16, page 3.7.16-1 Att 2, ITS B 3.7.16, page B 3.7.16-2 Att 5, NUREG 3.7.18, page 3.7-39 Att 5, NUREG B 3.7.18, page B 3.7-90 34

Spent Fuel Assembly Storage 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage

~Al .

~.1. lfo' 3 LCO 3.7.16 The combination of initial enrichment and burnup of each spent fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.

e..1+~r APPLICABILITY: Whenever any fuel assembly is stored in Region II ofAthe spent fuel pool~ north tilt pit.

or fnt.

ACTIONS

~------------------NOTE--------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME

A. Requirements of the LCO not met.

A.1 Initiate action to move the noncomplying fuel assembly from Region I I.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial Prior to enrichment and burnup of the fuel assembly storing the is in accordance with Table 3.7.16-1. fuel assembly in Region II Palisades Nuclear Plant 3.7.16-1 Amendment No. 01/20/98

!3~

Spent Fuel Assembly Storage B 3.7.16 BASES LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Table 3.7.16-1, in the accompanying LCO, ensures that the keff of the spent fuel pool will always remain< 0.95 assuming the pool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Table 3.7.16-1, in the accompanying LCO. Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Specification 4.3.1.1.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II of the spent fuel pool ~north tilt pit.

ti'.1r'ntr or the..

ACTIONS Required Action A.l is modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

When the configuration of fuel assemblies stored in Region II the spent fuel pool is not in accordance with Table 3.7.16-1, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Table 3.7.16-1.

SURVEILLANCE SR 3. 7 .16. 1 ,

REQUIREMENTS mx\ ~ 1.H,..\

Palisades Nuclear Plant B 3.7.16-2 01/20/98 J4-b

Spent Fuel Assembly Stora~~

Q) 3.7.~

APPLICABILITY : l(

x ACTIONS (fj .

CONDITION ' ...........

REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met.

A. l

~

l --------NOTE---------

LCO 3.0.3 is not applicable.

I ---------------------

Initiate action to Irrrnediately move the noncomplying

'{Region~

1I.

SURVEILLANCE REQUIREMENTS SURVEILLANCE . FREQUENCY

/{p

3. 7.~.l Vtrify by administrative means the initial Prior to enrichment and burnup ~he fuel assembly storing the i~ in accordance with

3.7.~l ~ fuel assembly

.E eclficatitn 4.3.ItU. ( ~ in ;rRegion ~~

]I:

x CEOG STS 3.7-39 Rev 1, 04/07,.95

Spent Fuel BASES (continued)

APPLICABILITY ACTIONS Required Action

is modified by i Note indiciting thit LCO 3.0.3 does not i pl

mov ng rra itt ue assem w e .

LCO 3.0.3 would not specify any action. If moving irradtited fuel assemblies while in HOOE 1, 2, 3, or 4, the fuel rDOvement ts 1ndependtnt of reactor operitton.

Therefore, tn either case, inibiltty to move fuel 1ss1mblits is not sufficient reason to require i reictor shutdown *

SURVEILLANCE REQUIREMENTS

@)

Cf) -r~le.

REFERENCES None.

f r1 ~ ft; pla..vlf\ d

.fv~) ~fi'. hfj I() C..

RA t ~ 17./to-/

( *I.~St:.Rl>

CEOG STS B 3.7-90 Rev 1, 04/07/95

- .\,

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.7, PLANT SYSTEMS NRC REQUEST:

3.7.17 Fuel Storage Pool Boron Concentration 3.7.17-1 CTS 4.2, Table 4.2.1, Item #7 ITS SR 3.7.17.1 DOC L.1 Comment: (Contractor comment 3.7.17-3) The removal of this CTS requirement appears acceptable; however, the DOC L.1 explains this CTS change but does not provide a specific technical justification for why this CTS requirement can be deleted. Provide this missing justification in a revision to the DOC.

Consumers Energv Response:

DOC L.1 has been revised to provide additional justification for the deletion of CTS 4.2, Table 4.2.1, Item #7.

Affected Submittal Pages:

Att 3, DOC 3.7.17, page 2 of 2 Att 4, NSHC 3.7.17, page 1 of 2 35

ATTAC1'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY A.5 CTS 3. l.5c requires that with specific activity of the secondary coolant

>0.1 µCi/gram DOSE EQUIVALENT 1-131, the plant must be placed in COLD SHUTDOWN. In proposed ITS the term is replaced with MODE 5 (see DOC A.4).

In proposed ITS 3.7.17 Applicability, the Specification is applicable in MODES 1, 2, 3, and 4. Placing the plant in COLD SHUTDOWN in CTS and having the Applicability in MODES 1, 2, 3, and 4 in proposed ITS is basically the same. This change is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M.l CTS 4.2 Table 4.2.1, item 7a, requires the specific activity of the secondary coolant system to be determined once per 31 days whenever the gross activity determination indicates iodine concentrations greater than 103 of the allowable limit, and once per 6 months whenever the gross activity determination indicates iodine concentrations below 10 3 of the allowable limit. Proposed ITS SR 3. 7 .17 .1 will require the specific activity to be determined once per 31 days. The proposed ITS SR will not contain the allowance to extend the SR interval to 6 months whenever the gross activity determination indicates iodine concentration below 103 of the allowable limit. This change does not adversely affect safety because the 31 day interval ensures that the specific activity is checked frequently enough to establish a trend to identify secondary activity problems in a timely manner. Deleting an allowance to extend an SR interval constitutes a more restrictive change. This change is consistent with NUREG-1432.

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

There were no "Removal of Details" associated with this specification.

LESS RESTRICTIVE CHANGES (L)

L.1 TS 4.2, Table 4.2.7 re ires a sample of secondary ca lant be analyzed for gross radioactivity 3 times ev ry 7 days with a maximum of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between samples. T is RA/ requirement has been eleted. The CTS contains no CO, limiting value, or Requi ed

~.7.tl* I Actions associated w. th this requirement in CTS, o y that sampling is required.

change is consider Less Restrictive ~ecause this ampling* requirement is delete .

This change is co istent with NUREG-1432.

Palisades Nuclear Plant Page 2of2 01/20/98 35 -Cl....

INSERT CTS 4.2, Table 4.2.1 requires a sample of secondary coolant to be analyzed for gross radioactivity 3 times every 7 days with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples. The CTS contains no LCO, limiting value, or Required Actions for secondary coolant gross radioactivity, only that sampling is required. The intent of this surveillance is to monitor the iodine concentration in the secondary coolant in order to determine the frequency at which an isotopic analysis for Dose Equivalent I-131 concentration in the secondary coolant is performed. The CTS requires an isotopic analysis for Dose equivalent I-131 of the secondary coolant once per 31 days whenever the gross activity indicates iodine concentrations greater than 10% of the allowable limit or, once per 6 months whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. -However as discussed in DOC M.1 for this specification, the extended surveillance interval of 6 months for the determination of Dose Equivalent I-131 in the secondary coolant has been proposed for deletion and that future testing be performed every 31 days.

Thus, the need to perform sampling of the secondary coolant for gross radioactivity is no longer necessary and has been delete in the ITS. This change is acceptable since gross radioactivity in the secondary coolant is not

- evaluated for radiological consequences in any of the accidents assumed in the FSAR, and the concentration of the Dose Equivalent I-131 in the secondary coolant will continue to be determined at an appropriate frequency. In addition, radiation monitoring instrumentation, controlled in accordance with the Offsite Dose Calculation Manual (e.g., SG blowdown monitors and condenser off gas monitor), is available to monitor increases in the radioactivity levels in the secondary coolant.* This change is consistent with NUREG-1432

35-b

LESS RESTRICTIVE CHANGE L.1 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY CTS 4.2, Table 4.2. 7 require a sample of secondary coolant be analyze for gross radioactivity 3 times every 7 ays with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months betwee samples. This RAI requirement has been delete . The CTS contains no LCO, limiting va e, or Required Actio 3.1.17- / associated with this requir ent in CTS, only that sampling is requir d. This change is considered Less Restricti e because this sampling requirement is de ted. This change is consistent with NURE 143

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 deletes the sample requirement for gross radioactivity of the secondary coolant. This sample does not have a detrimental impact on the integrity of any plant structure, system, or component. Deletion of this sample requirement will not alter the operation of any plant equipment, or otherwise increase its failure probability. As such, the probability of occurrence for a previously analyzed accident is not significantly increased.

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. Gross radioactivity of the secondary coolant is not an initial condition input assumed for any analyzed event.

The amount of Dose Equivalent 1-131 in the secondary coolant is the assumed parameter. The limit requirement for Dose Equivalent 1-131 remains unchanged and the sampling requirement has become more restrictive (see M. l). The deletion of the gross radioactivity sampling requirement does not affect the assumptions of an analyzed event. This change does not affect the performance of any credited equipment since the sample requirement is for an unassumed parameter. 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 1of2 01/20/98 35-C--

INSERT CTS 4.2, Table 4.2.1 requires a sample of secondary coolant to be analyzed for gross radioactivity 3 times every 7 days with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples. The CTS contains no LCO, limiting value, or Required Actions for secondary coolant gross radioactivity, only that sampling is required. The intent of this surveillance is to monitor the iodine concentration in the secondary coolant in order to determine the frequency at which an isotopic analysis for Dose Equivalent I-131 concentration in the secondary coolant is performed. The CTS requires an isotopic analysis for Dose equivalent I-131 of the secondary coolant once per 31 days whenever the gross activity indicates iodine concentrations greater than 10% of the allowable limit or, once per 6 months whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. However as discussed in DOC M.1 for this specification, the extended surveillance interval of 6 months for the determination of Dose Equivalent I-131 in the secondary coolant has been proposed for deletion and that future testing be performed every 31 days.

Thus, the need to perform sampling of the secondary coolant for gross radioactivity is no longer *necessary and has been delete in the ITS. This change is acceptable since gross radioactivity in the secondary coolant is not evaluated for radiological consequences in any of the accidents assumed in the FSAR, and the concentration of the Dose Equivalent I-131 in the secondary coolant will continue to be determined at an appropriate frequency. In addition, radiation monitoring instrumentation, controlled in accordance with the Offsite Dose Calculation Manual (e.g., SG blowdown monitors and condenser off gas monitor), is available to monitor increases in the radioactivity levels in the secondary coolant. This change is consistent with NUREG-1432.

3 s-d

ENCLOSURE 2 CONSUMERS ENERGY COMPANY PALISADES PLANT

- DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION EDITORIAL CHANGES

UHS 3.7.9-3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCO 3.7.9 The UHS shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 5bg, 2.5 SR 3.7.9.1 Verify water level of UHS is ~ §71.9 ft 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months above mean sea level.

SR 3.7.9.2 Verify water temperature of UHS is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

_s; 81.5°F.

Palisades Nuclear Plant 3.7.9-1 Amendment No. 01/20/98

___ I

AFW System B 3.7.5

B 3.7 PLANT SYSTEMS B 3.7.5 Auxiliary Feedwater (AFW) System BASES BACKGROUND The AFW System automatically supplies feedwater to the steam generators to remove decay heat from the Primary Coolant System upon the loss of normal feedwater supply. The AFW pumps take suction through a common suction line from the Condensate Storage Tank (CST) (LCO 3.7.6, "Condensate Storage and Supply") and pump to the steam generator secondary side v*ia two separate and independent flow paths to a common AFW supply header for each steam generator. The steam generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere from the steam generators via the Main Steam Safety Valves (MSSVs) (LCO 3.7.1, "Main Steam Safety Valves (MSSVs)") or Atmospheric Dump Valves (ADVs) (LCO 3.7.4, "Atmospheric Dump Valves (ADVs)"). If the main condenser is available, steam may be released via the turbine bypass valve .

The AFW System consists of two motor driven AFW pumps and one steam turbine driven pump configured into two trains.

One train (A/B) consists of a motor driven pump (P-8A) and the turbine driven pump (P-8B) in parallel, the discharges join together to form a common discharge. The A/B train common discharge separates to form two flow paths~hich cJ

~

feed each steam generator via each steam generate~ AFW X penetration. The second motor driven pump (P-8C) feeds both steam generators through separate flow paths via each steam .

generator AFW penetration and forms the other train (C).

The two trains join together at each AFW penetration to form a common supply to the steam generators. Each AFW pump is capable of providing 100% of the required capacity to the steam generators as assumed in the accident analysis. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system.

Each motor driven AFW pump is powered from an independent Class IE power supply, and feeds both steam generators .

Palisades Nuclear Plant B 3.7.5-1 01/20/98

AFW System B 3.7.5

BASES BACKGROUND The steam turbine driven AFW pump receives steam from either (continued) main steam header upstream of the Main Steam Isolation Valve (MSIV). Each of the steam feed lines will supply 100% of the requirements of the turbine driven AFW pump. The steam supply from steam generator E-50A receives anopen signal from the Auxiliary Feedwater Actuation Signal (AFAS) instrumentation. The steam supply from steam generator E-50B does not. This steam source is a manual backup. The turbine driven AFW pump feeds both steam generators through the same flow paths as motor driven AFW pump P-8A.

One pump at full flow is sufficient to remove decay heat and cool the plant to Shutdown Cooling (SOC) System entry conditions.

The AFW System supplies feedwater to the steam generators during normal plant startup, shutdown, and hot standby conditions.

The AFW System is designed to supply sufficient water to the steam generator(s) to remove decay heat with steam generator pressure at the setpoint of the MSSVs, with exception of AFW pump P-8C. If AF~Wu p P-8C is used, operator action may be required to eithe rip two of four Primary Coolant Pumps (PCPs), start an ditional AFW pump, or reduce steam generator pressure. This will allow the required flowrates to the steam generators that are assumed in the safety analyses. Subsequently, the AFW System supplies sufficient water to cool the plant to SOC entry conditions, and steam is released through the ADVs, or the turbine bypass valve if the condenser is available.

The AFW System actuates automatically on low steam generator level by an AFAS as described in LCD 3.3.3, Engineered 11 Safety Feature (ESF) Instrumentation and 3.3.4, ESF 11 11 Logic. The AFAS initiates signals for starting the AFW 11 pumps and repositioning the valves to initiate AFW flow to the steam generators. The actual pump starts are on an as 11 required basis. P-8A is started initially, if the pump 11 fails to start, or if the required flow is not established in a specified period of time, P-8C is started. If P-8A and P-8C do not start, or if required flow is not established in a specified period of time, then P-8B is started.

The AFW System is discussed in the FSAR, Section 9.7

(Ref. 1).

Palisades Nuclear Plant B 3.7.5-2 01/20/98

AFW System B 3.7.5

BASES APPLICABLE The AFW System mitigates the consequences of any event with SAFETY ANALYSES a loss of normal feedwater.

The design basis of the AFW System*is to supply water to the steam generator to remove decay heat and other residual ~

heat, by delivering at least the minimum required flow rate to the steam generators at pressures corres ondin i+t. ~

lowest MSSV set pressure plus 3% wit exception of AFW pump P-8C. If AFW pump P-8C is used, operator action maybe ~

required to either trip two of the four PCPs, start"an additional AFW pump or reduce steam generator pressure. This will allow the required flowrate to the steam generators that are assumed in the safety analyses.

The limiting Design Basis Accident for the AFW System is a loss of normal feedwater.

In addition, the minimum available AFW flow and system characteristics are serious considerations in the analysis of a small break loss of coolant accident.

The AFW System design is such that it can perform its function following loss of normal feedwate~combined with a loss of offsite power with one AFW pump injecting AFW to one steam generator.

The AFW System satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO This LCO requires that two AFW trains be OPERABLE to ensure that the AFW System will perform the design safety function to mitigate the consequences of accidents that could result in overpressurization of the primary coolant pressure boundary. Three independent AFW pumps, in two diverse trains, ensure availability of residual heat removal capability for all events accompanied by a loss of offsite power and a single failure. This is accomplished by powering two pumps from independent emergency buses. The third AFW pump is powered by a diverse means, a steam driven turbine supplied with steam from a source not isolated by the closure of the MSIVs .

Palisades Nuclear Plant B 3.7.5-3 01/20/98

~-

Condensate Storage and Supply B 3.7.6 BASES APPLICABLE The Condensate Storage and Supply provides condensate to SAFETY ANALYSES remove decay heat and to cool down the plant following all events in the accident analysis, discussed in the FSAR, Chapters 5 and 14. For anticipated operational occurrences and accidents which do not affect the OPERABILITY of the steam generators, the analysis assumption is generally 30 minutes at MODE 3, steaming through the MSSVs followed by a cooldown to Shutdown Cooling (SOC) entry conditions at the design cooldown rate.

The Condensate Storage and Supply satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO To satisfy accident analysis assumptions, the CST and T-81 must contain sufficient cooling water to remove decay heat for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following a* reactor trip from 102%. RTP. This*

amount of time allows for cool down of the PCS to SOC entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this the CST and T-81 must retain sufficient water to ensure adequate net positive suction head for the* AFW pumps, and makeup for steaming required to remove decay heat.

The combined CST and T-81 level required is a usable volume of at least 100,000 gallons, which is based on holding the plant in MODE 3 for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , followed by a cooldown to SOC entry conditions at approximately 75°F per hour. This basis Cl

{).)as ~ established by the Systematic Evaluation Program. ')(

OPERABILITY of the Condensate Storage and Supply System is determined by maintaining the combined tank levels at or above the minimum required volume.

APPLICABILITY In MODES 1, 2, and 3, and in MODE 4, when steam generator is being relied upon for heat removal, the Condensate Storage and Supply is required to be OPERABLE.

In MODES 5 and 6, the Condensate Storage and Supply is not required because the AFW System is not required .

Palisades Nuclear Plant B 3.7.6-2 01/20/98

CCW System B 3.7.7

B 3.7 PLANT SYSTEMS B 3.7.7 Component Cooling Water (CCW) System BASES BACKGROUND The CCW System provides a heat sink for the removal of process and operating heat from safety related components

,during a Design Basis Accident (OBA) or transient. During normal operation, the CCW System also provides this function for various nonessential components, as well as the spent fuel pool. The CCW System serves as a ba~rier to the release of radioactive byproducts between potentially radioactive systems and the Service Water System (SWS), and thus to the environment.

The CCW System consists of three pumps connected in parallel to common suction and discharge headers. The discharge cJ.nJ

- K header splits into two parallel heat exchangers-lthen combines again into a common di stri buti on header ~~ [wA1c1i heat l cads. A common surge tank provides the necessary net Sv,,llcs positive suction head for the CCW pumps and a surge volume for the system. A train of CCW~~ be that equipment electrically connected to a common safety bus necessary to transfer heat acquired from the various heat loads to the SWS. There are two CCW trains, each associated with a Safeguards Electrical Distribution Train which are described in Specification 3.8.9, "Distribution Systems - Operating."

The CCW train associated with the Left Safeguards Electrical Distribution Train consists of two CCW pumps (P-52A, P-52C),

both CCW heat exchangers (E-54A, E-54B), the CCW surge tank (T-3), associated piping, valves, and controls for the equipment to perform their safety function. The CCW train associated with the Right Safeguards Electrical Distribution Train consists of one CCW pump (P-528), both CCW heat exchangers (E-54A, E-548), the CCW surge tank (T-3),

associated piping, valves, and controls for the equipment to perform their safety function. The pumps and valves are automatically started upon receipt of a safety injection actuation signal and all essential valves are aligned to their post accident positions. CCW valve repositioning also occurs following a Recirculation Actuation Signal (RAS) which aligns associated valves to provide full cooling to the CCW heat exchangers during the recirculation phase following a design basis Loss of Coolant Accident (LOCA).

Palisades Nuclear Plant B 3.7.7-1 01/20/98

S\~S B 3.7.8 BASES SURVEILLANCE SR 3.7.8.l (continued)

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

SR 3.7.8.2 This SR verifies proper automatic operation of the SWS valves on an actual or simulated actuation signal. Specific signals (e.g., safety injection) are tested under Section 3.3, 11 Instrumentation. 11 If the isolation valve for the noncritical service water header CV-1359) or for

. containment air cooler VHX-4 iso at on (CV-0869) fail to du<. \-6 t~ diVvSi~*A ~ close, then both trains of SWS are considered inoperable.sz......i 6-f c..~ 1 ~ wa;:cu, .flow ~This Surveillance is not required for valves that are

'1 d .

locked, sealed, or otherwise secured in the required

position under administrative controls. This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, 11 Instrumentation, 11 the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

SR J. 7.8.3 .

The SR verifies proper automatic operation of the SWS pumps on an actual or simulated actuation signal in the 11 with standby power available 11 mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, 11 Electrical Power Systems. 11 This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, 11 Instrumentation, 11 the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

Palisades Nuclear Plant B 3.7.8-5 01/20/98

UHS B 3.7.9

B 3.7 PLANT SYSTEMS B 3.7.9 Ultimate Heat Sink (UHS)

BASES BACKGROUND The UHS provides a heat sink for process and operating heat from safety related components during a Design Basis Accident (OBA) or transient, as well as during normal operation. This is done utilizing the Service Water System (SWS).

The UHS has been defined as Lake Michigan. The two principal functions of the UHS are the dissipation of residual heat after reactor shutdown, and dissipation of residual heat after an accident.

The basic performance requirements are that an adequate Net Positive Suction Head (NPSH) to the SWS pumps be available, and that the design basis temperatures of safety related equipment not be exceeded.

Additional information on the design and operation of the system along with a list of components served can be found in FSAR, Section 9.1 (Ref. 1).

APPLICABLE The UHS is the sink for heat removed from the reactor core SAFETY ANALYSES following all accidents and anticipated operational occurrences in which the plant is cooled down and placed on shutdown cooling. Maximum post accident heat load occurs between 20 to 40 minutes after a design basis Loss of Coolant Accident (LOCA). Near this time, the plant switches from injection to recirculation, and the containment cooling J'GGH systems are required to remove the core decay heat. ~ c~

TuSCfl.T H~e e13eFatiF1~ liAlits ue based on conservative heat transfer ccJ analyses for the worst case LOCA. FSARmiS ction 14.18 (Ref. 2) and Design Basis Document (DBD 1.02 (Ref. 3) (

provid~the details of the analysis wh1 forms the basis

~the operating limits. The assumptions include: worst )(

expected meteorological conditions, conservative uncertainties when calculating decay heat, and the worst case single active failure .

The UHS satisfies Criterion 3 of 10 CFR 50.36(c)(2) .

Palisades Nuclear Plant B 3.7.9-1 01/20/98

The minimum water level of the UHS is based on the NPSH requirements for the SWS pumps.

The NPSH calculation assumes a minimum water level of 4 feet above the bottom of the pump suction bell which corresponds to an elevation of 557.25 ft. Violation of the SWS pump submergence requirement should never become a factor unless the Lake Michigan water level falls below the top of the sluice gate opening which is at elevation 568.25 ft. Early warning of a falling intake water level is provided by the intake structure level alann. The nominal lake level is approximately 580 ft mean sea level. The minimum water temperature of the UHS is ...

UHS B 3.7.9

BASES LCO The UHS is required to be OPERABLE. The UHS is considered OPERABLE if it contains a sufficient volume of water at or below the maximum temperature that would allow the SWS to operate without the loss of NPSH, and without exceeding the maximum design temperature of the equipment served by the SWS. To meet this condition, the UHS temperature should not exceed 8l.5°F and the level should not fall below~ ft above mean sea level during normal plant operation.~ S~&.2.5 APPLICABILITY In MODES 1, 2, 3, and 4, the U~S is a normally operating system that is required to support the OPERABILITY* of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODES 5 and 6, the OPERABILITY requirements of the UHS are determined by the systems it supports.

ACTIONS A.1 and A.2 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 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 3.7.9.1 REQUIREMENTS This SR verifies adequate cooling can be maintained. The level specified also ensures sufficient NPSH is available for operating the SWS pumps. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to the trending of the parameter variations during the applicable MODES. This SR verifies that the UHS water level is ~ §71.~ ft above mean x sea level as measured within the boundaries1of the intake structure.

S"CP~.is Palisades Nuclear Plant B 3.7.9-2 01/20/98

SFP Level B 3.7.14

B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool (SFP) Water Level BASES BACKGROUND The minimum water level in the SFP meets the assumptions of iodine decontamination factors following a fuel handling or cask drop accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the SFP design is given in the FSAR, Section 9.11 (Ref. 1), and the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.4 (Ref. 2). The assumptions of fuel handling and fuel cask drop accidents are given in the FSAR, Section 14.19 and 14.11 (Refs. 3 and 4), respectiv.ely.

APPLICABLE The minimum water level in the SFP meets the assumptions SAFETY ANALYSES of fuel handling or fuel cask drop accident analyses described in References 3 and 4 and are consistent with the assumptions of Regul ato.ry Gui de 1. 25 (Ref. 5). The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to a person at the exclusion area boundary is well within the 10 CFR 100 (Ref. 6) limits.

f'O.Oool'll<>O A~prrliR§ te Reference ~here is 23 ft of water between the top of the damaged ruel assembly and the fuel pool x surface for a fuel handling or fuel cask drop accident.

This LCD preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single assembly, dropped and lying horizontally on top of the spent fuel racks, there may be < 23 ft of water above the top of the assembly and the ~urface, by the width of the assembly. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rods fail from a hypothetical maximum drop.

The SFP water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2) .

Palisades Nuclear Plant B 3.7.14-1 01/20/98

SFP Boron Concentration B 3.7.15 BASES APPLICABILITY This LCO applies whenever fuel assemblies are stored in the*

spent fuel pool until a complete spent fuel pool verification of the stored assemblies has been performed following the last movement of fuel assemblies in the spent fuel pool. This LCO does not apply following the verification since the verification would confirm that there a~e no misloaded fuel assemblies. With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.

ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, th~

fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

A.l. A.2.1. and A.2.2 When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. This does not preclude the movement of fuel assemblies to a safe position. In addition, action must be immediately initiated to restore boron concent~ati to within limit. Alte.!!1ptely, beginning a verificatio f the SFP fuel location~to ensure proper locations of e fuel can be performed. .

Palisades Nuclear Plant B 3.7.15-2 01/20/98

Spent Fuel Assembly Storage B 3.7.16

B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage BASES

BACKGROUND The spent fuel storage facility is designed to store either new (nonirradiated) nuclear fuel assemblies, or used (irradiated) fuel assemblies in a vertical configuration underwater. The storage pool is sized to store 892 irradiated fuel assemblies, which includes storage for failed fuel canisters. The spent fuel storage racks are .. Cd grouped into two regi ens, Region I and Region I I perr (Seis""', '(

Figure 3.7.16-1. The racks are designed as a~ I iC1t-k~ory' structure able to withstand seismic events. Region I contains racks in the spent fuel pool having a 10.25 inch center-to-center spacing and a single rack in the north tilt pit having a 11.25 inch by 10.69 inch center-to-center spacing. Region II contains racks in both the spent fuel pool and the north tilt pit having a 9.17 inch center-to-center spacing. Because of the smaller spacing and poison concentration, Region II racks have more limitations for fuel storage th~n Region I racks. Further information on these limitations can be found in Section 4.0, Design Features. These limitations 11 11 (e.g., enrichment, burnup) are sufficient to maintain a keff of ~ 0.95 for spent fuel of ori in

of u to 4.40%. IHOwever, 1g er initial nrichment fuel crSSeiTI6lies ares ored in the spen fuel pool, the must be stored in a ch kerboard patter taking into ac unt fuel

£&

~

burnup to mai tain a keff of 0. or less.

APPLICABLE The spent fuel storage facility is designed for SAFETY ANALYSES noncriticality by use of adequate spacing, and flux trap" 11 construction whereby the fuel assemblies are inserted into neutron absorbing stainless steel cans.

The spent fuel assembly storage satisfies Criterion 2 of 10 CFR 50.36(c)(2) .

Palisades Nuclear Plant B 3.7.16-1 01/20/98

Secondary Specific Activity B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Secondary Specific Activity BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Primary Coolant System (PCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives, and thus is indication of current conditions. During transients, I-131 spikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

This limit is lower than the activity value that might be expected from a 1 gpm tube leak of primary coolant at the

limit of 1.0 µCi/gm as assumed in the safety analyses with exception of the control rod ejection analysis which assumes 0.6 gpm. LCO 3.4.13, "PCS Operational LEAKAGE," is more restrictive in that the limit for a primary to secondary tube leak is 0.3 gpm. The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and primary coolant LEAKAGE. Most of the iodine isotopes have short half lives (i.e.,< 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ).

Operating a plant at the allowable limits ~o~ld result in a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Exclusion Area Boundary (EAB) exposure well within the 10 CFR 100 (Ref. 1) limits.

Palisades Nuclear Plant B 3.7.17-1 01/20/98

Secondary Specific Activity B 3.7.17

BASES APPLICABILITY In MODES 1, 2, 3, and 4, the limits on secondary specific activity apply due to the potential for secondary steam releases to the atmosphere.

In MODES 5 and 6, the steam generators are not being used for heat removal. Both the PCS and steam generators are at low pressure or depressurized, and primary to secondary LEAKAGE is minimal. Therefore, monitoring of secondary specific activity is not required.

ACTIONS A.1 and A.2 DOSE EQUIVALENT wrI-131 exceeding the allowable value in the secondary coolan is an indication of a problem in the PC~

and contributes increased post accident doses. If secondary specific activity cannot be restored to within limits in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. *To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in 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 3.7.17.1 REQUIREMENTS This SR ensures that the secondary specific activity is within the limits of the accident analysis. A gamma isotope analysis of the secondary coolant, which determines DOSE EQUIVALENT I-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in primary coolant activity or LEAKAGE. The 31 day Frequency is based on the detection of increasing trends of the level of DOSE EQUIVALENT I-131, and allows for appropriate action to be taken to maintain levels below the LCO limit.

Palisades Nuclear Plant B 3.7.17-3 01/20/98

.4. 2

The Control Room Ventilation and Isolation System (andAhe flJelAto/agd 1Area}!EP~har;?oa1Z£xh;l!jst/Sysfemlshall be demonstrated to be OPERABLE by the f'OTTowing tests: ( ~?:--0

,7.1?

@ I 2. /At l?ast/§nce/per/ref}?el ifg c#le Ay:I

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

a.

SR 3.1. 10. "3 b.

(1t1iD 5~ 3.7.io.r>§ sl-( fr~D SR~-1.lo.3 Nott? (@) L./ (

4-14 Amendment No. 8-l-, ~. ~

ATTACHMENT.3 DISCUSSION OF CHANGES SPECIFICATION 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM ADMINISTRATIVE CHANGES (A)

A. l All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.

Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.

A.2 CTS 3.4.2 and 3,4.3 require that if~o!onent(s) listed in Speeification 3.4.1 is cc/_

X inoperable for more than the time sp citied, the plant must be placed in HOT SHUTDOWN. In proposed ITS 3. . Required Action B.l, the CTS term is replaced with MODE 3. This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

A.3 CTS 3.4.4 specifies that valves, interlocks and piping that are directly associated with the "above" (CTS 3.4.1) components shall meet the same requirements as listed for that component. CTS 3 .4.5 specifies that valves, interlocks and piping which is associated with the containment cooling system and not covered by CTS 3 .4.4 may be inoperable for no more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if it is required to function during an accident. These requirements are addressed by the definition of OPERABILITY which requires that all associated equipment be OPERABLE. In the proposed ITS, all equipment in a particular train which is required to function during an accident must be OPERABLE and all equipment in the train will have the same Completion Time. This is an administrative change since the requirement remains that all equipment in a train of containment cooling must be OPERABLE. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 1of5 01/20/98

A.4 SPEC

.l ~-4*~,

CATION 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM ATTACHMENT 3 DISCUSSION OF CHANGES CTS 3.3.2 and 3.4.3 require that with the Required Action and associated Completion .

Time not met the plant must be placed in COLD SHUTDOWN. In proposed f.cl X

ITS 3.7.7 Required Action B.2, the CTS term is replaced with MODE 5. This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

A.5 CTS 3.4.3 states " .... Continued power operation with one component out of service shall be as specified in Section 3 .4.2, with the permissible period in inoperability starting at the time that the first of the two components became inoperable." This explanatory information on the usage rules of technical specifications is addressed in the proposed ITS Section 1. 3, "Completion Times," and does not need to be addressed cJ in the Actions of proposed ITS 3. 7'.Trhis is considered to be an administrative X _...

change since the requirements on complying with the completion times is addressed in the proposed ITS. This change is consistent with NUREG-1432.

A.6 The Note added to proposed SR 3.7.7.1 to aid the operator in the prevention of entering an inappropriate LCO. The Note reminds the operator that loss of CCW flow to a component may render that component inoperable but does not affect the I OPERABILITY of the CCW System. This change is considered administrative that this .I is a clarifier to the operator to prevent confusion. This change is consistent with NUREG-1432.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M.1 CTS 3.3.1, 3.3.2, 3.4.1, and 3.4.2 establish the Applicability for the various components which comprise the CCW by stating that "the reactor shall not be made critical. ... unless all of the following conditions are met." The Applicability of the CCW in proposed ITS 3.7.7 is MODES 1, 2, 3, and 4. As such, the requirements associated with CTS 3.3.1, 3.3.2, 3.4.1, and 3.4.2 have been revised to be more restrictive by requiring the CCW to also be OPERABLE during the additional MODES 3 and 4. SRs 3.7.7.2 and 3.7.7.3 are modified by a Note which states that these SRs are not required to be met in MODE 4. This is due to the instrumentation providing the signals are not required in MODE 4. This change keeps consistency with ITS 3.3.3, "ESF Instrumentation," and current licensing basis. This change is an additional restriction on plant operations and is consistent with NUREG-1432.

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

l

LESS RESTRICTIVE CHANGES (L)

ATTACH1\1ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.10, CRV FILTRATION L.1 CTS 4.2, Table 4.2.3, item~f requires a verification that the Control Room Ventilation.system automatically switches into the emergency* mode of operation on a "containment high pressure and high radiation test signal." The Applicability of this requirement is "above COLD SHUTDOWN, during REFUELING OPERATIONS, during movement of irradiated fuel assemblies, and during movement of a fuel cask in or over the Spent Fuel Pool. " Proposed SR 3. 7 .10. 3 requires a verification that each CRV Filtration train actuates on an actual or simulated actuatio~ signal. The requirement and Applicability of CTS 4.2, Table 4.2.3, item<>>ais similar to the requirement and Applicability of SR 3.7.10.3. However, SR 3.7.10.3 is further modified by a Note which states that the SR is "not required to be met during movement of irradiate fuel assemblies in the SFP, or during movement of a fuel cask in or over the SFP." The purpose of this Note is to exclude the requirement of the SR during those plant evolutions in which no instrumentation is available to actuate the CRV System. The CRV System is designed to automatically switch to the emergency mode of operation on a "containment high pressure or _containment high radiation signal. " The instruments used to initiate these actuation signals are not capable of detecting an increase in radiation levels in the fuel handling building, and as such, can not provide automatic actuation of the CRV System in the event of a fuel handling accident or cask drop accident in the SFP. Therefore,..-the addition of the Note in SR 3. 7 .10. 3 establishes consistency with the design of the CRV System and the requirement of the SR. During movement of irradiate fuel assemblies in the SFP, or during movement of a fuel cask in or over the SFP, manual operator action is necessary to initiate the emergency filtration mode of the CRV System.¢

Palisades Nuclear Plant Page 4of4 01/20/98

ENCLOSURE 3 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS

RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7, PLANT SYSTEMS Page Change Instruct;ons 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 3.7.9-1 ITS 3.7.9-1 03/15/99 Tech change ITS 3.7.12-1 ITS 3. 7 .12-1 03/15/99 RA! 3.7.12-1 ITS 3.7.12-2 ITS 3.7.12-2 03/15/99 RA! 3.7.12-1 ITS 3.7.12-3 ------------ 03/15/99 RA! 3.7.12-1 ITS 3.7.13-1 ITS 3.7.13-1 03/15/99 RA! 3. 7 .13-2 ITS 3.7.16-1 ITS 3.7.16-1 03/15/99 RA! 3.7.16-3 ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL ITS B 3.7.2-2 ITS B 3.7.2-2 03/15/99 RA! 3.7.2-2 ITS B 3.7.2-3 ITS B 3.7.2-3 03/15/99 RAI 3.7.2-2 ITS B 3.7.2-4 ITS B 3.7.2-4 03/15/99 RA! 3.7.2-2

- ITS B 3.7.2-5 ITS B 3.7.3-1 ITS B 3.7.3-2 ITS B 3.7.3-3 ITS B 3.7.2-5 ITS B 3.7.3-1 ITS B 3.7.3-2 ITS B 3.7.3-3 03/15/99 03/1~/99 03/15/99 03/15/99 RA! 3.7.2-2 RA! 3.7.3-1 RA! 3.7.3-2 RA! 3.7.3-2 RA! 3.7.3-5 RA! 3.7.3-6 ITS B 3.7.3-4 ITS B 3.7.3-4 03/15/99 RA! 3.7.3-4 ITS B 3.7.3-5 ITS B 3.7.3-5 03/15/99 RA! 3.7.3-4 ITS B 3.7.5-1 ITS B 3. 7. 5-1 03/15/99 editorial ITS B 3.7.5-2 ITS B 3.7.5-2 03/15/99 editorial ITS B 3.7.5-3 ITS B 3.7.5-3 03/15/99 editorial ITS B 3. 7. 6-2 ITS B 3. 7. 6-2 03/15/99 editorial ITS B 3.7.6-3 ITS B 3.7.6-3 03/15/99 RA! 3.7.6-1 ITS B 3.7.7-1 ITS B 3.7.7-1 03/15/99 editorial ITS B 3.7.8-5 ITS B 3. 7. 8-5 03/15/99 editorial ITS B 3.7.9-1 ITS B 3.7.9-1 03/15/99 Tech change ITS B 3.7.9-2 ITS B 3.7.9-2 03/15/99 Tech change ITS B 3.7.9-3 ITS B 3. 7. 9-3 03/15/99 Tech change ITS B 3.7.12-1 ITS B 3.7.12-1 03/15/99 RA! 3.7.12-1 ITS B 3.7.12-2 ITS B 3.7.12-2 03/15/99 RA! 3.7.12-1 ITS B 3.7.12-3 ITS B 3.7.12-3 03/15/99 RA! 3.7.12-1 ITS B 3.7.12-4 ITS B 3.7.12-4 03/15/99 RA! 3.7.12-1 ITS B 3.7.12-5 ITS B 3.7.12-5 03/15/99 RAI 3.7.12-1 ITS B 3.7.12-6 ITS B 3.7.12-6 03/15/99 RAI 3.7.12-1

-------------- ITS B 3.7.12-7 03/15/99 RAI 3.7.12-1 1

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7, PLANT SYSTEMS REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT#

ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL (continued)

ITS B 3.7.13-2 ITS B 3. 7 .13-2 03/15/99 RAI 3. 7 .13-2 ITS B 3.7.14-1 ITS B 3.7.14-1 03/15/99 editorial ITS B 3. 7 .15-2 ITS B 3.7.15-2 03/15/99 editorial ITS B 3. 7 .16-1 ITS B 3.7.16-1 03/15/99 editorial ITS B 3.7.16-2 ITS B 3.7.16-2 03/15/99 RAI 3.7.16-1 RAI 3.7.16-3 ITS B 3.7.16-3 -------------- 03/15/99 RAI 3.7.16-3 ITS B 3.7.17-1 ITS B 3.7.17-1 03/15/99 editorial ITS B 3.7.17-3 ITS B 3.7.17-3 03/15/99 editorial ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL CTS 3.7.5, pg 3-38a CTS 3.7.5, pg 3-38a 03/15/99 RAI 3.7.5-1 CTS 3.7.7, pg 3-29a CTS 3.7.7, pg 3-29a 03/15/99 RAI 3.7.7-1 CTS 3.7.10, pg 4-14 CTS 3.7.10, pg 4-14 03/15/99 editorial CTS 3.7.12, pg 3-47 CTS 3.7.12, pg 3-47 03/15/99 RAI 3.7.12-1 CTS 3.7.12, pg 3-46 CTS 3.7.12, pg 3-46 03/15/99 RAI 3.7.12-1 CTS 3.7.12, pg 4-14 CTS 3.7.12, pg 4-14 03/15/99 RAI 3.7.12-1

- DOC 3.7.5, pg 2 of 7 DOC 3.7.7, pg 1 of 5 DOC 3.7.7, pg 2 of 5 DOC 3.7.10, pg 4 of 4 DOC 3.7.12, pg 1 of 4 DOC 3.7.12, pg 2 of 4 DOC 3.7.5, pg 2 of 7 DOC 3.7.7, pg 1 of 5 DOC 3.7.7, pg 2 of 5 DOC 3.7.10, pg 4 of 4 DOC 3.7.12, pg 1 of 3 DOC 3.7.12, pg 2 of 3 03/15/99 .

03/15/99 03/15/99 03/15/99 03/15/99 03/15/99 RAI 3.7.5-2 editorial editorial editorial RAI 3.7.12-1 RAI 3.7.12-1 DOC 3.7.12, pg 3 of 4 DOC 3.7.12, pg 3 of 3 *03/15/99 RAI 3.7.12-1 DOC 3.7.12, pg 4 of 4 --------------------- 03/15/99 RAI 3.7.12-1 DOC 3.7.17, pg 1 of 2 DOC 3.7.17, pg 1 of 3 03/15/99 RAI 3.7.17-1 DOC 3.7.17, pg 2 of 2 DOC 3.7.17, pg 2 of 3 03/15/99 RAI 3.7.17-1

DOC 3.7.17, pg 3 of 3 03/15/99 RAI 3.7.17-1 ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL NSHC 3.7.12, pg 1 of 2 NSHC 3.7.12, pg 1 of 2 03/15/99 RAI 3.7.12-1 NSHC 3.7.12, pg 2 of 2 NSHC 3.J.12, pg 2 of 2 03/15/99 RAI 3.7.12-1 NSHC 3.7.17, pg 1 of 2 NSHC 3.7.17, pg 1 of 2 03/15/99 RAI 3.7.17-1 NSHC 3.7.17, pg 2 of 2 NSHC 3.7.17, pg 2 of 2 03/15/99 RAI 3.7.17-1 ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG 3.7-21 NUREG 3.7-21 03/15/99 Tech change NUREG 3.7-29 NUREG 3.7-29 03/15/99 RAI 3.7.13-2 NUREG 3.7-31 NUREG 3.7-31 03/15/99 RAI 3.7.12-1 NUREG 3.7-31 ~insert ------------ 03/15/99 RAI 3.7.12-1 NUREG 3.7-31 inserts NUREG 3.7-31 inserts 03/15/99 RAI 3.7.12-1

NUREG 3.7-32 NUREG 3.7-33 NUREG 3.7-39 NUREG 3.7-32 NUREG 3.7-33 NUREG 3.7-39 2

03/15/99 03/15/99 03/15/99 RAI 3.7.12-1 RAI 3.7.12-1 RAI 3.7.16-1 RAI 3.7.16-3

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.7, PLANT SYSTEMS REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT#

ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG B 3.7-7 NUREG B 3.7-7 03/15/99 RAI 3.7.2-2 NUREG B 3.7-7 insert NUREG B 3.7-7 insert 03/15/99 RAI 3.7.2-2 NUREG B 3.7-8 NUREG B 3.7-8 03/15/99 editorial

NUREG B 3.7-8 insert 03/15/99 editorial NUREG B 3.7-13 insert NUREG B 3.7-13 insert 03/15/99 RAI 3.7.3-1 RAI 3.7.3-2 NUREG B 3.7-14 NUREG B 3.7-14 03/15/99 RAI 3.7.3-2 RAI 3.7.3-5 NUREG B 3.7-14 insert NUREG B 3.l-14 insert 03/15/99 RAI 3.7.3-2 NUREG B 3.7-15 NUREG B 3.7-15 03/15/99 RAI 3.7.3-6 NUREG B 3.7-17 NUREG B 3.7-17 03/15/99 RAI 3.7.3-4 NUREG B 3.7-34 NUREG B 3.7-34 03/15/99 RAI 3.7.6-1 NUREG B 3.7-36 insert NUREG B 3.7-36 insert 03/15/99 - RAI 3.7 .6-1 NUREG B 3.7-44 insert NUREG B 3.7-44 insert 03/15/99 RAI 3.7.6-1 NUREG B 3.7-47 NUREG B 3.7-47 03/15/99 Tech change

NUREG B 3.7-47 insert 03/15/99 Tech change NUREG B 3.7-49 NUREG B 3.7-49 03/15/99 Tech change

NUREG B 3.7-65 insert 03/15/99 RAI 3.7.13-1 NUREG B 3.7-67 NUREG B 3.7-67 03/15/99 RAI 3. 7.13-2 NUREG B 3.7-67 insert NUREG B 3.7-67 insert 03/15/99 RAI 3. 7 .13-2 NUREG B 3. 7-71 . NUREG B 3.7-71 03/15/99 RAI 3.7.12-1 NUREG B 3.7-71 insert NUREG B 3.7-71 insert 03/15/99 RAI 3.7.12-1 NUREG B 3.7-72 NUREG B 3.7-72 03/15/99 RAI 3.7.12-1 NUREG B 3.7-72 in. (2pgs) NUREG B 3.7-72 in. (3pgs) 03/15/99 RAI 3.7.12-1 NUREG B 3.7-73 NUREG B 3.7-73 03/15/99 RAI 3.7.12-1 NUREG B 3.7-73 insert NUREG B 3.7-73 insert 03/15/99 RAI 3.7.12-1 NUREG B 3.7-74 NUREG B 3.7-74 03/15/99 RAI 3.7.12-1 NUREG B 3.7-74 insert -------------- 03/15/99 RAI 3.7.12-1 NUREG B 3.7-75 NUREG B 3.7-75 03/15/99 RAI 3.7.12-1 NUREG B 3.7-75 insert NUREG B 3.7-75 insert 03/15/99 RAI 3.7.12-1 NUREG B 3.7-76 NUREG B 3.7-76 03/15/99 RAI 3.7.12-1 NUREG B 3.7-89 NUREG B 3.7-89 03/15/99 editorial NUREG B 3.7-90 NUREG B 3.7-90 03/15/99 RAI 3.7.16-1 RAI 3.7.16-3 ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL JFD 3.7.14, pg 1 of 2 JFD 3.7.14, pg 1 of 4 03/15/99 RAI 3.7.12-1 JFD 3.7.14, pg 2 of 2 JFD 3.7.14, pg 2 of 4 03/15/99 RAI 3.7.12-1

JFD 3.7.14, pg 3 of 4 03/15/99 RAI 3.7.12-1

JFD 3.7.14, pg 4 of 4 03/15/99 RAI 3.7.12-1 JFD 3.7.18, pg 1 of 2 JFD 3.7.18, pg 1 of 2 03/15/99 RAI 3.7.16-1 JFD 3.7.18, pg 2 of 2 JFD 3.7.18, pg 2 of 2 03/15/99 RAI 3.7.16-1 3

UHS 3.7.9

3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCD 3.7.9 The UHS shall be OPERABLE.

APPLICABILITY: MOqES l, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify. water level of UHS is ~ 568.25 ft 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months above meari sea level.

SR 3.7.9.2 Verify water temperature of UHS is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

81.5°F.

Palisades Nuclear Plant 3.7.9-1 Amendment No. 03/15/99

Fuel Handling Area Ventilation System 3.7.12

3.7 PLANT SYSTEMS 3.7.12 Fuel Handling Area Ventilation System LCO 3.7.12 The Fuel Handling Area Ventilation System shall be OPERABLE with one fuel handling area exhaust fan aligned to the emergency filter bank and in operation.

APPLICABILITY: During movement of irradiated fuel assemblies in the fuel handling building when irradiated fuel assemblies with

< 30 days decay time are in the fuel handling building, During movement of a fuel cask in or over the SFP when irradiated fuel assemblies with< 90 days decay time are in the fuel handling building, During CORE ALTERATIONS when irradiated fuel assemblies with

< 30 days decay time are in the containment with the equipment hatch open, During movement of irradiated fuel assemblies in the containment when irradiated fuel assemblies with

< 30 days decay time are in the containment'with the equipment hatch open.

ACTIONS CONDITION .REQUIRED ACTION COMPLETION TIME A. Fuel Handling Area A.1 Suspend movement of Immediately Ventilation System not fuel assemblies.

aligned or in-operation. AND OR A.2 Suspend CORE Immediately ALTERATIONS.

Fuel Handling Area Ventilation System AND inoperable.

A.3 Suspend movement of a Immediately fuel cask in or over the SFP.

Palisades Nuclear Plant 3.7.12-1 Amendment No. 03/15/99

~------

Fuel Handling Area Ventilation System 3.7.12

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Perform required Fuel Handling Area In accordance Ventilation System filter testing in with the accordance with the Ventilation Filter Ventilation Testing Program. Filter Testing Program SR 3.7.12.2 Verify the flow rate of the Fuel Handling 18 months Area Ventilation System, when aligned to the emergency filter bank, is ~ 5840 cfm and ~ 8760 cfm .

Palisades Nuclear Plant 3~7.12-2 Amendment No. 03/15/99

ESRV Dampers

3. 7 .13

3.7 PLANT SYSTEMS 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers LCO 3.7.13 Two ESRV Damper trains shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ESRV A.1 Initiate action to Immediately Damper trains isolate associated inoperable. ESRV Damper train(s).

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify each ESRV Damper train closes on an 31 days actual or simulated actuation signal.

Palisades Nuclear Plant 3.7.13-1 Amendment No. 03/15/99

Spent Fuel Assembly Storage 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage LCO 3.7.16 The combination of initial enrichment and burnup of each spent fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.

APPLICABILITY: Whenever any fuel assembly is stored in Region II of either the spent fuel pool or the north tilt pit.

ACTIONS

NOTE--------------------------------------

LCO 3.0.3 is not ap~licable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Initiate action to Immediately LCO not met. move the noncomplying fuel assembly from Region I I.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial Prior to enrichment and burnup of the fuel assembly storing the is in accordance with Table 3.7.16-1. fuel assembly in Region II Palisades Nuclear Plant 3.7.16-1 Amendment No. 03/15/99

MS I Vs B 3.7.2

BASES APPLICABLE The design basis of the MSIVs is established by the SAFETY ANALYSES containment analysis for the Main Steam Line Break (MSLB) inside containment, as discussed in the FSAR, Section 14.18 (Ref. 2). It is also influenced by the accident analysis of the MSLB events presented in the FSAR, Section 14.14 (Ref. 3). The MSIVs are swing disc check valves. The inherent characteristic of this type of valve allows for reverse flow through the valve on a differential pressure even if the valve is closed. In the event of an MSLB, if the MSIV associated with the unaffected steam generator fails to close, both steam generators may blowdown. This failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a double steam generator blowdown event, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences.

Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial improvement to plant safety.

There are three different limiting MSLB cases that have been evaluated, one for fuel integrity and two for containment analysis (one for containment temperature and one for containment pressure). The limiting case for containment temperature is the hot full power MSLB inside containment following a turbine trip. At hot full power, the stored energy in the primary coolant is maximized.

The limiting case for the containment analysis for containment pressure and fuel integrity is the hot zero power MSLB inside containment. At zero power, the steam generator inventory and temperature are at their maximum, maximizing the analyzed mass and energy release to the containment. Reverse flow due to the open MSIV bypass valves, contributes to the total release of the additional mass and energy. With the most reactive control rod assumed stuck in the fully withdrawn position, there is an increased possibility that the core will return to power. The core is ultimately shut down_by a combination of doppler feedback, steam generator dryout, and borated water injection delivered by the Emergency Core Cooling System.

Palisades Nuclear Plant B 3.7.2-2 03/15/99

MS I Vs B 3.7.2 BASES APPLICABLE The accident analysis compares several different MSLB events SAFETY ANALYSES against different acceptance criteria. The MSLB outside (continued) containment upstream of the MSIV is limiting for offsite dose, although a break in this short section of main steam header has a very low probability. The MSLB inside containment at hot full power is the limiting case for a post trip return to power. The analysis includes scenarios with offsite power available and with a loss of offsite power following a turbine trip.

With offsite power available, the primary coolant pumps continue to circulate coolant through the steam generators, maximizing the Primary Coolant System (PCS) cooldown. With a loss of offsite power, the response of mitigating systems, such as the High Pressure Safety Injection (HPSI) pumps, is delayed.

The MSIVs serve only a safety function and remain open during power operation. These valves operate under the following situations:

a. An MSLB inside containment. For this accident scenario, steam is discharged into containment from both steam generators until closure of the MSIV in the intact steam generator occurs. After MSIV closure, steam is discharged into containment only from the affected steam generator.

b. A break outside of containment and upstream from the MSIVs. This scenario is not a containment pressurization concern. The uncontrolled blowdown of more than one steam generator must be prevented to limit the potential for uncontrolled PCS cooldown and positive reactivity addition. Closure of the MSIVs limits the blowdown to a single steam generator.

c. A break downstream of the MSIVs. This type of break will be .isolated by the closure of the MSIVs. Events such as increased steam flow through the turbine or the turbine bypass valve will also terminate on closure of the MSIVs.

d. A steam generator tube rupture. For this scenario, closure of the MSIVs isolates the affected steam generator from ~he intact steam generator and minimizes radiological releases.

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

Palisades Nuclear Plant B 3.7.2-3 03/15/99

MS I Vs B 3.7.2 BASES LCO This LCO requires that the MSIV in each of the two steam lines be OPERABLE. The MSIVs are considered OPERABLE when the isolation times are within limits, and they close on an isolation signal.

This LCO provides assurance that the MSIVs will perform their design safety function to mitigate the consequences of accidents that could result in offsite exposures comparable to the 10 CFR 100.11 (Ref. 4) limits or the NRC staff approved licensing basis.

APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3 except when both MSIVs are closed and deactivated when there is significant mass and energy in the PCS and steam generators. When the MSIVs are closed, they are already performing their safety function. Deactivation can be accomplished by the removal of the motive force (e.g., air) to the valve to prevent valve opening.

In MODE 4, the steam generator energy is low; therefore, the MSIVs are not required to be OPERABLE.

In MODES 5 and 6, the steam generators do not contain much energy because their temperature is below the boiling point of water; therefore, the MSIVs are not required for isolation of potential high energy secondary system pipe breaks in these MODES.

ACTIONS With one MSIV inoperable in MODE 1, time is allowed to restore the component to OPERABLE status. Some repairs can be made to the MSIV with the plant hot. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable, considering the probability of an accident occurring during* the time p~riod that would require closure of the MSIVs.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is greater than that normally allowed for containment isolation valves because the MSIVs are valves that isolate a closed system penetrating containment. -

Palisades Nuclear Plant B 3.7.2-4 03/15/99

MS I Vs B 3.7.2 BASES ACTIONS B.1 (continued)

If the MSIV cannot be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Condition C would be entered. The Completion Time is reasonable, based on operatin~ experience, to reach MODE 2 in an orderly manner and without challenging plant systems.

C.l and C.2 Condition C is modified by a Note indicating that separate Condition entry is allowed for each MSIV.

Since the MSIVs are required to be OPERABLE in MODES 2 and 3, the inoperable MSIVs may either be restored to

OPERABLE status or closed. When closed, the MSIVs are already in the position required by the assumptions in the

- safety analysis .

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is consistent with that allowed in Condition A.

Inoperable MSIVs that cannot be restored to OPERABLE status within the specified Completion Time, but are closed, must be verified on a periodic basis to be closed. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering judgment, MSIV status indications available in the control room, and other administrative controls, to ensure these valves are in the closed position.

D.1 and D.2 If the MSIVs cannot be restored to OPERABLE status, or closed, within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply.

To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach-the required plant conditions from MODE 2 in an orderly manner and without challenging plant systems.

Palisades Nuclear Plant B 3.7.2-5 03/15/99

MFRVs and MFRV Bypass Valves B 3.7.3 B 3.7 PLANT SYSTEMS B 3.7.3 Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves BASES BACKGROUND The MFRVs and MFRV bypass valves in conjunction with feed pump speed, control Main Feedwater (MFW) flow to the steam generators for level control during normal plant operation.

The valves also isolate MFW flow to the secondary side of the steam generators following a High Energy Line Break (HELB). Closure of the MFRVs and MFRV bypass valves terminates flow to both steam generators. Closure of the MFRV and MFRV bypass valve effectively terminates the addition of feedwater to an affected steam generator, limiting the mass and energy release for Main Steam Line Breaks (MSLBs) inside containment, and reducing the cooldown effects.

The MFRVs and MFRV bypass valves isolate MFW in the event of a secondary side pipe rupture inside containment to limit the quantity of high energy fluid that enters containment through the break. Controlled addition of Auxiliary Feedwater (AFW) is provided by a separate piping system.

One MFRV and one MFRV bypass valve are located on each MFW line outside containment. The piping volume from the valves to the steam generator must be accounted for in calculating mass and energy releases following an MSLB.

The MFRVs and MFRV bypass valves close on receipt of a isolation signal generated by either; steam generator low pressure from its respective steam generator, or containment high pressure. These isolation signals also actuate the Main Steam Isolation Valves (MSIVs) to close. The MFRVs and MFRV bypass valves may also be actuated manually. The MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping that fail "as-is" on a loss of air. If required, MFW isolation can be accomplished using manually operated valves upstream or downstream of the MFRVs and MFRV Bypass valves. In addition, each MRFV is equipped with a handwheel that can be used to isolate this MFW flowpath.

A description of the MFRVs and MFRV bypass valves is found in the FSAR, Section 10.2.3 (Ref. 1) .

Palisades Nuclear Plant B 3.7.3-1 03/15/99

MFRVs and MFRV Bypass Valves B 3.7.3 BASES APPLICABLE Closure of the MFRVs is an assumption in the MSLB SAFETY ANALYSES containment response analysis. Closure of the MFRVs and MFRV bypass valves is also assumed in the MSLB core response (DNB) analysis.

Failure of an MFRV or MFRV bypass valve to close following a MSlB can result in additional mass and energy to the steam generators contributing to cooldown. This failure also results in additional mass and energy releases following an MSLB event. However, this failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a single steam generator blowdown event with continued feedwater, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial improvement to plant safety.

The MFRVs and MFRV bypass valves satisfy Criterion 3 of 10 CFR 50.36(c)(2).

LCO This LCO ensures that the MFRVs and MFRV bypass valves will isolate MFW flow to the steam generators following an MSLB.

This* LCO requires that both MFRVs and both MFRV bypass valves be OPERABLE. The MFRVs and MFRV bypass valves are considered OPERABLE when the isolation times are within limits, and are closed on an isolation signal.

Failure to meet the LCO requirements can result in additional mass and energy being released to containment following an MSLB inside containment .

Palisades Nuclear Plant B 3.7.3-2 03/15/99

MFRVs and MFRV Bypass Valves B 3.7.3

BASES APPLICABILITY All MFRVs and MFRV bypass valves must be OPERABLE, or either closed and deactivated, or isolated by closed manually actuated valves, whenever there is significant mass and energy in the Primary Coolant System and steam generators.

In MODES 1, 2, and 3, the MFRVs or MFRV bypass valves are required to be OPERABLE, except when both MFRVs and both MFRV bypass valves are either closed and deactivated, or isolated by closed manually actuated valves, in order to limit the amount of available fluid that could be added to containment in the case of a secondary system pipe break inside containment. When the valves are either closed and deactivated, or isolated by closed manually actuated valves, they are already performing their safety function.

Once the valves are closed, deactivation can be accomplished by the removal of the motive force (e.g., electrical power, air) to the valve to prevent valve opening. Closing another manual valve in the flow path either remotely (i.e., control room hand switch) or locally by manual operation satisfies isolation requirements .

In MODES 4, 5, and 6, steam generator energy is low.

Therefore, the MFRVs and MFRV bypass valves are not required to be OPERABLE.

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each valve.

A.1 and A.2 With one MFRV or MFRV bypass valve inoperable, action must be taken to close or isolate the inoperable valve(s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . When these valve(s) are closed or isolated, they are performing their required safety function (e.g., to isolate the line).

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable to close the MFRV or MFRV bypass valve, which includes performing a controlled plant shutdown to condition that supports isolation of the affected valve(s). .

Palisades Nuclear Plant B 3.7.3-3 03/15/99

MFRVs and MFRV Bypass Valves B 3.7.3 BASES ACTIONS B.1 and B.2 (continued)

If the MFRVs or MFRV bypass valves cannot be restored to OPERABLE status, closed, or isol~ted in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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 3.7.3.1 REQUIREMENTS This SR verifies the closure time for each MFRV and MFRV bypass valve is ~ 22.0 seconds on an actual or simulated actuation signal. Specific signals (e.g., steam generator low pressure and containment high pressure) are tested under Section 3.3, "Instrumentation." The MFRV and MFRV bypass valves closure times are bounding values assumed in the MSLB containment response and core response (DNB) analyses (Refs. 3 and 4). This SR is normally performed upon returning the plant to operation following a refueling outage. The MFRVs and MFRV bypass valves should not be tested at power since even a part stroke exercise increases the risk of a valve closure with the plant generating power.

As these valves are not stroke tested at power, they are exempt from the ASME Code,Section XI (Ref. 2) requirements during operation in MODES 1 and Z.

The Frequency is 18 months. The 18 month Frequency for valve closure time is based on the refueling cycle.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency.

Palisades Nuclear Plant B 3.7.3-4 03/15/99

MFRVs and MFRV Bypass Valves B 3.7.3

BASES REFERENCES 1. FSAR, Section 10.2.3

2. ASME, Boiler and Pressure Vessel Code,Section XI, Inservice Inspection, Article IWV-3400

3. FSAR, Section 14.18.2

4. FSAR, Section 14.14

Palisades Nuclear Plant B 3.7.3-5 03/15/99

AFW System B 3.7.5 B 3.7 PLANT SYSTEMS B 3.7.5 Auxiliary Feedwater (AFW) System BASES BACKGROUND The AFW System automatically supplies feedwater to the steam generators to remove decay heat from the Primary Coolant System upon the loss of normal feedwater supply. The AFW pumps take suction through a common suction line from the Condensate Storage Tank (CST) (LCO 3.7.6, "Condensate Storage and Supply") and pump to the steam generator secondary side via two separate and independent flow paths to a common AFW supply header for eath steam generator. The steam generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere from the steam generators via the Main Steam Safety Valves (MSSVs) (LCO 3.7 .1, "Main Steam Safety Valves (MSSVs) or Atmospheric Dump Valves (ADVs) (LCO 3.7.4, 11

)

"Atmospheric Dump Valves (ADVs) 11

). If the main condenser is available, steam may be released via the turbine bypass valve.

The AFW System consists of two motor driven AFW pumps and one steam turbine driven pump configured into two trains.

One train (A/B) consists of a motor driven pump (P-8A) and the turbine driven pump (P-8B) in parallel, the discharges join together to form a common discharge. The A/B train common discharge separates to form two flow paths, which feed each steam generator via each steam generator s AFW 1

penetration. The second motor driven pump (P-8C) feeds both steam generators through separate flow paths via each steam generator AFW penetration and forms the other train (C).

The two trains join together at each AFW penetration to form a common supply to the steam generators. Each AFW pump is capable of providing 100% of the ~equired capacity to the steam generators as assumed in the accident analysis. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system.

Each motor driven AFW pump is powered from an independent Class lE power supply, and feeds both steam generators.

Palisades Nuclear Plant B 3.7.5-1 03/15/99

AFW System B 3.7.5

BASES BACKGROUND The steam turbine driven AFW pump receives steam from either (continued) main steam header upstream of the Main Steam Isolation Valve (MSIV). Each of the steam feed lines will supply 100% of the requirements of the turbine driven AFW pump. The steam supply from steam generator E-50A receives an open signal from the Auxi 1i ary Feedwater Actuation Si gna 1 (AFAS) instrumentation. The steam supply from steam generator E-50B does not. This steam source is a manual backup. The turbine driven AFW pump feeds both steam generators through the same flow paths as motor driven AFW pump P-BA.

One pump at full flow is sufficient to remove decay heat and cool the plant to Shutdown Cooling (SOC) System entry conditions.

The AFW System supplies feedwater to the steam generators during normal plant startup, shutdown, and hot standby conditions.

The AFW System is designed to supply sufficient water to the steam generator(s) to remove decay heat with steam generator pressure at the setpoint of the MSSVs, with exception of AFW pump P-BC. If AFW pump P-BC is used, operator action may be required to either trip two of four Primary Coolant Pumps (PCPs), start an additional AFW pump, or reduce steam generator pressure. This will allow the required flowrates to the steam generators that are assumed in the safety analyses. Subsequently, the AFW System supplies sufficient water to cool the plant to SOC entry conditions, and steam is released through the ADVs, or the turbine bypass valve if the condenser is available.

The AFW System actuates automatically on low steam generator level by an AFAS as described in LCD 3.3.3, "Engineered Safety Feature (ESF) Instrumentation" and 3.3.4, "ESF Logic." The AFAS initiates signals for starting the AFW pumps and repositioning the valves to initiate AFW flow to the steam generators. The actual pump starts are on an "as required" basis. P-BA is started initially, if the pump fails to start, or if the required flow is not established in a specified period of time, P-BC is started. If P-BA and P-BC do not start, or if required flow is not established in a specified period of time, then P-BB is started.

The AFW System is discussed in the FSAR, Section 9.7

(Ref. 1).

Palisades Nuclear Plant B 3.7.5-2 03/15/99

AFW System B 3.7.5

BASES APPLICABLE The AFW System mitigates the consequences of any event with SAFETY ANALYSES a loss of normal feedwater.

The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat, by delivering at least the minimum required flow rate to the steam generators at pressures corresponding to the lowest MSSV set pressure plus 3% with the exception of AFW pump P-8C. If AFW pump P-8C is used, operator action may be required to either trip two of the four PCPs, start an additional AFW pump or reduce steam generator pressure. This will allow the required flowrate to the steam generators that are assumed in the safety analyses.

The limiting Design Basis Accident for the AFW System is a loss of normal feedwater.

In addition, the minimum available AFW flow and system characteristics are serious considerations in the analysis of a small break loss of coolant accident .

The AFW System design is such that it can perform its function following loss of normal feedwater combined with a loss of offsite power with one AFW pum~ injecting AFW to one steam generator.

The AFW System satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO This LCO requires that two AFW trains be OPERABLE to ensure that the AFW System wi 11 ,perform the design safety function to mitigate the consequences of accidents that could result in overpressurization of the primary coolant pressure boundary. Three independent AFW pumps, in two diverse trains, ensure availability of residual heat removal capability for all events accompanied by a loss of offsite power and a single failure. This is accomplished by powering two pumps from independent emergency buses. The third AFW pump is powered by a diverse means, a steam driven turbine supplied with steam from a source not isolated by the closure of the MSIVs.

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Condensate Storage and Supply B 3.7.6

BASES APPLICABLE The Condensate Storage and Supply provides condensate to SAFETY ANALYSES remove decay heat and to cool down the plant following all events in the accident analysis, discussed in the FSAR, Chapters 5 and 14. For anticipated operational occurrences and accidents which do not affect the OPERABILITY of the steam generators, the analysis assumption is generally 30 minutes at MODE 3, steaming through the MSSVs followed by a cooldown to Shutdown Cooling (SOC) entry conditions at the design cooldown rate.

The Condensate Storage and Supply satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO To satisfy accident analysis assumptions, the CST and T-81 must contain sufficient cooling water to remove decay heat for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following a reactor trip from 102% RTP. This amount of time allows for cool down of the PCS to SOC entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this the CST and T-81 must retain sufficient water to ensure adequate net positive suction head for the AFW pumps, and makeup for steaming required to remove decay heat.

The combined CST and T-81 level required is a usable volume of at least 100,000 gallons, which is based on holding the plant in MODE 3 for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , followed by a cooldown to SOC entry conditions at approximately 75°F per hour. This basis was established by the Systematic Evaluation Program.

OPERABILITY of the Condensate Storage and Supply System is determined by maintaining the combined tank levels at or above the minimum required volume.

APPLICABILITY In MODES 1, 2, and 3, and in MODE 4, when steam generator is being relied upon for heat removal, the Condensate Storage and Supply is required to be OPERABLE.

In MODES 5 and 6, the Condensate Storage and Supply is not required because the AFW System is not required.

Palisades Nuclear Plant B 3.7.6-2 03/15/99

Condensate Storage and Supply B 3.7.6

BASES ACTIONS A.1 and A.2 If the condensate volume is not within the limit, the OPERABILITY of the backup water supplies must be verified by administrative means within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

OPERABILITY of the backup feedwater supplies must include verification of the OPERABILITY of flow paths from the Fire Water System and SWS to the AFW pumps, and availability of the water in the backup supplies. The Condensate Storage and Supply volume must be returned to OPERABLE status within 7 days, as the backup supplies may be performing this function in addition to their normal functions. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable, based on operating experience, to verify the OPERABILITY of the Fire Water System and SWS. Additionally, verifying the backup water supplies every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is adequate to ensure the backup water supplies continue to be available. The 7 day

Completion Time is reasonable, based on OPERABLE backup water supplies being available, and the low probability of an event requiring the use of the water from the CST and T-81 occurring during this period.

B.1 and B.2 If the condensate volume cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4, without reliance on steam generator for heat removal, within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-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 .

Palisades Nuclear Plant B 3.7.6-3 03/15/99

CCW System B 3.7.7

B 3.7 PLANT SYSTEMS B 3.7.7 Component Cooling Water (CCW) System BASES BACKGROUND The CCW System provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (OBA) or transient. During normal operation, the CCW System also provides this function for various nonessential components, as well as the spent fuel pool. The CCW System serves as a barrier to the release of radioactive byproducts between potentially radioactive systems and the Service Water System (SWS), and thus to the environment.

The CCW System consists of three pumps connected in parallel to common suction and discharge headers. The discharge header splits into two parallel heat exchangers and then combines again into a common distribution header which supplies various heat loads. A common surge tank provides the necessary net positive suction head for the CCW pumps and a surge volume for the system. A train of CCW is considered to be that equipment electrically connected to a common safety bus necessary to transfer heat acquired from the various heat loads to the SWS. There are two CCW trains, each associated with a Safeguards Electrical Distribution Train which are described in Specification 3.8.9, "Distribution Systems - Operating." The CCW train associated with the Left Safeguards Electrical Distribution Train consists of two CCW pumps (P-52A, P-52C), both CCW heat exchangers (E-54A, E-54B), the CCW surge tank (T-3),

associated piping, valves, and controls for the equipment to perform their safety function. The CCW train associated with the Right Safeguards Electrical Distribution Train consists of one CCW pump (P-52B); both CCW heat exchangers (E-54A, E-54B), the CCW surge tank (T-3), associated piping, valves, and controls for the equipment to perform their safety function. The pumps and valves are automatically started upon receipt of a safety injection actuation signal and all essential valves are aligned to their post accident positions. CCW valve repositioning also occurs following a Recirculation Actuation Signal (RAS) which aligns associated valves to provide full cooling to the CCW heat exchangers during the recirculation phase following a design basis Loss of Coolant Accident (LOCA).

Palisades Nuclear Plant B 3.7.7-1 03/15/99

sws B 3.7.8

BASES SURVEILLANCE SR 3.7.8.1 (continued)

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

SR 3.7.8.2 This SR verifies proper automatic operation of the SWS valves on an actual or simulated actuation signal. Specific signals (e.g., safety injection) are tested under Section- 3.3, "Instrumentation." If the isolation valve for the noncritical service water header (CV-1359) or for containment air cooler VHX-4 (CV-0869) fail to close, then both trains of SWS are considered inoperable due to the diversion of cooling water flow. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required,position under administrative controls. This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met -in this MODE. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

SR 3:7.8.3 The SR verifies proper automatic operation of the SWS pumps on an actual or simulated actuation signal in the "with standby power available" mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, "Electrical Power Systems." This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Sur~eillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

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UHS B 3.7.9 B 3.7 PLANT SYSTEMS B 3.7.9 Ultimate Heat Sink (UHS)

BASES BACKGROUND The UHS provides a heat sink for process and operating heat from safety related components during a Design Basis Accident (OBA) or transient, as well as during normal operation. This is done utiliiing the Service Water System (SWS).

The UHS has been defined as Lake Michigan. The two principal functions of the UHS are the dissipation of residual heat after reactor shutdown, and dissipation of residual heat after an accident.

The basic performance requirements are that an adequate Net Positive Suction Head (NPSH) to the SWS pumps be available, and that the design basis temperatures of safety related equipment not be exceeded.

Additional information on the design and operation of the system along with a list of components served can be found in FSAR, Section 9.1 (Ref. 1).

APPLICABLE The UHS is the sink for heat removed from the reactor core SAFETY ANALYSES following all accidents and anticipated operational occurrences in which the plant is cooled down and placed on shutdown cooling. Maximum post accident heat load occurs between 20 to 40 minutes after a design basis Loss of Coolant Accident (LOCA). Near this time, the plant switches from injection to recirculation, and the containment cooling systems are required to remove the core decay heat.

Palisades Nuclear Plant B 3.7.9-1 03/15/99

UHS B 3.7.9

BASES APPLICABLE The minimum water level of the UHS is based on the NPSH SAFETY ANALYSES requirements for the SWS pumps. The NPSH calculation*

(continued) assumes a minimum water level of 4 feet above the bottom of the pump suction bell which corresponds to an elevation of 557.25 ft. Violation of the SWS pump submergence requirement should never become a factor unless the Lake Michigan water level falls below the top of the sluice gate opening which is at elevation 568.25 ft. Early warning of a falling intake water level is provided by the intake structure level alarm. The nominal lake level is approximately 580 ft mean sea level. The minimum water temperature of the UHS is based on conservative heat transfer analyses for the worst case LOCA. FSAR, Section 14.18 (Ref. 2) and Design Basis Document (DBD) 1.02 (Ref. 3) provide the details of the analysis which forms the basis for these operating limits. The assumptions include: worst expected meteorological conditions, conservative uncertainties when calculating decay heat, and the worst case single active failure.

The UHS satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO The UHS is required to be OPERABLE. The UHS is considered OPERABLE if it contains a sufficient volume of water at or below the maximum temperature that would allow the SWS to operate without the loss of NPSH, and without exceeding the maximum design temperature of the equipment served by the SWS. To meet this condition, the UHS temperature should not exceed 81.5°F and the level should not fall below 568.25 ft above mean sea level during normal plant operation.

APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODES 5 and 6, the OPERABILITY requirements of the UHS are determined by the systems it supports.

Palisades Nuclear Plant B 3.7.9-2 03/15/99

UHS B 3.7.9 BASES ACTIONS A.1 and A.2 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 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 3.7.9.1 REQUIREMENTS This SR verifies adequate cooling can be maintained. The level specified also ensures sufficient NPSH is available for operating the SWS pumps. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to the trending of the parameter variations during the applicable MODES. This SR verifies that the UHS water level is ~ 568.25 ft above mean sea level as measured within the boundaries of the intake

structure.

SR 3.7.9.2 This SR verifies that the SWS is available to provide adequate cooling for the maximum accident or normal design heat loads following a OBA. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to the trending of the parameter variations during the applicable MODES. This SR verifies that the water temperature from the UHS is

81.5°F.

REFERENCES 1. FSAR, Section 9.1

2. FSAR, Section 14.18

3. Design Basis Document (DBD) 1.02, 11 Service Water System 11 Palisades Nuclear Plant B 3.7.9-3 03/15/99

Fuel Handling Area Ventilation System B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Fuel Handling Area Ventilation System BASES BACKGROUND The Fuel Handling Area Ventilation System filters airborne radioactive particulates from the area of the spent fuel pool following a fuel handling accident or a fuel cask drop accident. The fuel handling area is served by two separate subsystems one being part of the original plant design, and the other being added as part of the Auxiliary Building Addition.

The original plant design consists of a supply plenum and an exhaust plenum including associated ductwork, dampers, and instrumentation. The suoply plenum contains one prefilter, two heating coils, and one supply fan. The exhaust plenum contains two filter banks (normal and emergency) configured in a parallel flow arrangement, and two independent exhaust fans which draw air from a cormion duct. The "normal filter bank" contains a prefilter and a High Efficiency Particulate Air (HEPA) filter. The "emergency filter bank 11 contains a prefilter, HEPA filter, and an activated charcoal filter.

The Auxiliary Building Addition, which was added to serve the spaces at the north end of the spent fuel pool, also consist of a supply plenum and exhaust plenum. The supply plenum is configured similar to the supply plenum provided in the original plant design and includes one prefilter, two heating coils, and one supply fan. The exhaust plenum is diff~rent from the original plant design in that it only contains one filter bank consisting of a prefilter and HEPA filter, and two common exhaust .fans.

During normal plant operations, the Fuel Handling Area Ventilation System supplies filtered and heated (as needed) outside air to the fuel handling area. The exhaust fans draw air from the fuel handling area through the normally aligned prefilters and HEPA filters and discharge it to the unit stack by way of the main ventilation exhaust plenum.

Palisades Nuclear Plant B 3.7.12-1 . 03/15/99

Fuel Handling Area Ventilation System B 3.7.12 BASES BACKGROUND During plant evolutions when the possibility for a fuel (continued) handling accident or fuel cask drop accident exist, the Fuel Handling Area Ventilation System is configured such that all fans are stopped except one exhaust fan in the original plant subsystem aligned to the emergency filter bank. The 11 11 11 normal filter bank in the original plant design is 11 isolated by closing its associated inlet damper. Thus, in the event of a fuel handling accident, the fuel handling area atmosphere will be filtered for the removal of airborne fission products prior to being discharged to the outside environment.

The Fuel Handling Area Ventilation System is discussed in the FSAR, Sections 9.8, 14.11 and 14.19 (Refs. 1, 2, and 3) because it may be used for normal, as well as post accident, atmospheric cleanup functions.

APPLICABLE The Fuel Handling Area Ventilation System is designed to SAFETY ANALYSES mitigate the consequences of a fuel handling accident or fuel cask drop accident by limiting the amount of airborne radioactive material discharged to the outside atmosphere.

The results and major assumptions used in the analysis of the fuel handling accident are presented in FSAR Section 14.19. For the purpose of defining the upper limit of the radiological consequences of a fuel handling accident, it is assumed that a fuel bundle is dropped during fuel handling activities and all the fuel rods in the equivalent of an entire assembly (216) fail. The bounding fuel* handling accident is assumed to occur in containment two days after shutdown. No containment isolation is

assumed to occur. As such, the released fission products escape to the environment with no credit for filtration.

The re~ults of this analysis have shown that the offsite doses resulting from this event are within the guideline of 10 CFR 100. In the event a fuel handling accident were to occur in the fuel handling area, the radioactive release would pass through the emergency filter bank significantly 11 11 reducing the amount of radioactive material released to the environment. Thus, the consequences of a fuel handling

accident in the fuel handling area are deemed acceptable with or without th~_ emergency filter bank in operation 11 11 since they are no more severe than the consequences of a fuel handling accident in containment .

Palisades Nuclear Plant B 3.7.12-2 03/15/99

_J

Fuel Handling Area Ventilation System B 3.7.12 BASES APPLICABLE The results and major assumptions used in the analysis SAFETY ANALYSES of the fuel cask drop accident are presented in FSAR (continued) Section 14.11. For the purpose of defining the upper limit of the radiological consequences of a fuel cask drop accident, it is assumed that all 73 fuel assemblies in a 7 x 11 Westinghouse spent fuel pool rack with a minimum decay of 30 days are damaged and release their fuel rod gap inventories. Three fuel cask drop scenarios were analyzed to encompass all fuel cask drop events. They are:

1. A fuel cask drop onto 30 day decayed fuel with the Fuel Handling Area Ventilation System aligned for emergency filtration with a conservative amount of unfiltered leakage. All isolatable unfiltered leak path are assumed to be isolated prior to event initiation.

2. A fuel cask drop onto 30 day decayed fuel with the Fuel Handling Area Ventilation System aligned for emergency filtration with a conservative amount of unfiltered leakage. This scenario determined the

3.

maximum amount of non-isolatable unfiltered leakage than can exist and still meet offsite dose limits.

This scenario also assumes isolation of isolable leak paths prior to event initiation.

A fuel cask drop onto 90 day decayed fuel without the Fuel Handling Area Ventilation System aligned for emergency filtration. This scenario needs no assumptions as to unfiltered leakage or post-accident unfiltered leak path isolation times since all radiation is assumed to be released unfiltered from the fuel handling area.

The results of the analysis show that the radiological consequences of a fuel cask drop in the spent fuel pool meet the acceptance criteria of Regulatory Guide 1.25 (Ref. 4) and NUREG-0800 Section 15.7.5 (Ref. 5) for all scenarios.

In addition, the dose from all scenarios are less than 25%

of the dose guidelines in 10 CFR 100. For scenario 2, the analysis shows that a maximum of 20% charcoal filter bypass from non-isolatable leak paths can be acco1m1odated while still meeting 25% of the 10 CFR 100 guidelines .

Palisades Nuclear Plant B 3.7.12-3 03/15/99

Fuel Handling Area Ventilation System B 3.7.12 BASES APPLICABLE Filtration of the fuel handling area atmosphere following a SAFETY ANALYSES fuel handling accident is not necessary to maintain the (continued) offsite doses within the guidelines of 10 CFR 100. Thus, a total system failure would not impact the margin of safety as described in the safety analysis. However, analysis has shown that post accident filtration by the Fuel Handling Area Ventilation System provides significant reduction in offsite doses by limiting the release of airborne radioactivity. Therefore, for the fuel handling accident, the Fuel Handling Area Ventilation System satisfies Criterion 4 of 10 CFR 50.36(c)(2).

Filtration of the fuel handling area atmosphere following a fuel cask drop on irradiated fuel assemblies with < 90 days decay is required to maintain the offsite doses within the guidelines of 10 CFR 100. Therefore, for the fuel cask drop accident, the Fuel Handling Area Ventilation System satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO The LCO for the Fuel Handling Area Ventilation System ensures filtration of the fuel handling area atmosphere is immediately available in the event of a fuel handling accident, or a fuel cask drop accident. As such, the LCD requires the Fuel Handling Area Ventilation System to be OPERABLE with one fuel handling area exhaust fan aligned to the "emergency filter bank" and in operation.

The Fuel Handling Area Ventilation System is considered OPERABLE when the individual components necessary to control exposure in the fuel handling building are OPERABLE. The Fuel Handling Area Ventilation System is considered OPERABLE when:

a. One exhaust fan is aligned to the "emergency filter bank" and in operation to ensure the air discharged to the main ventilation exhaust plenum has been filtered.

Operation of only one fuel handling area exhaust fan ensures the design flow rate of the "emergency filter bank" is not exceeded.

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration functions; and

Palisades Nuclear Plant B 3.7.12-4 03/15/99

Fuel Handling Area Ventilation System B 3.7.12 BASES LCO c. Ductwork and dampers are OPERABLE, and air circulation (continued) can be maintained. Inclusive to the requirement to align the emergency filter bank is that the normal 11 11 11 filter bank is isolated by its associated inlet 11 damper to prevent the release of unfilted air.

APPLICABILITY The Fuel Handling Area Ventilation System must be Operable, aligned, and in operation whenever the potential exists for an accident that results in the release of radioactive material to the fuel handling area atmosphere that could exceed previously approved offsite dose limits if released unfiltered to the outside atmosphere. As such, the Fuel Handling Area Ventilation System is required; during movement of irradiated fuel assemblies in the fuel handling building when irradiated fuel assemblies with< 30 days decay time are in the fuel handling building; during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in containment when irradiated fuel assemblies with < 30 days decay time are in the containment with the equipment hatch open, and during movement of a fuel cask in or over the spent fuel pool when irradiated fuel assemblies with< 90 days are in the fuel handling building.

The requirement for the Fuel Handling Area Ventilation System does not apply during movement of irradiated fuel assemblies or CORE ALTERATIONS when all irradiated fuel assemblies in the fuel handling building, or all irradiated fuel assemblies in the containment with the equipment hatch open, have decayed for 30 days or greater since the dose consequences from a fuel handling accident would be of the same magnitude without the filters operating as the dose consequences would be with the filters operating and two days decay. In addition, the requirement for the Fuel Handling Area Ventilation System does not apply during fuel cask movement when all irradiated fuel assemblies in the fuel handling building have decayed 90 days or greater since the dose consequences remain less than 25% of the guidelines of 10 CFR 100 .

Palisades Nuclear Plant B 3.7.12-5 03/15/99

Fuel Handling Area Ventilation System B 3.7.12 BASES ACTIONS . A.1 and A.2 If the Fuel Handling Area Ventilation System is not aligned to the "emergency filter bank", or one exhaust fan is not in operation, or the system is inoperable for any reason, action must be taken to place the unit in a condition in which the LCO does not apply. Therefore, activities involving the movement of irradiated fuel assemblies, CORE ALTERATIONS, and movement of a fuel cask in or over the spent fuel pool, must be suspended immediately to minimize the potential for a fuel handling accident.

The suspension of fuel movement, CORE ALTERATIONS, and fuel cask movement shall not preclude the completion of placing a fuel assembly, core component, or fuel cask in a safe position.

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS This SR verifies the performance of Fuel Handling Area Ventilation System filter testing in accordance with the Ventilation Filter Testing Program. The Fuel Handling Area Ventilation System filter tests are in accordance with the Regulatory Guide 1.52 (Ref. 6) as described in Ventilation Filter Testing Program. The Ventilation Filter Testing Program includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and *following specific operations). Specific test frequencies and additional information are discussed in detail in the Ventilation Filter Testing Program.

Palisades Nuclear Plant B 3.7.12-6 03/15/99

Fuel Handling Area Ventilation System B 3.7.12 BASES SURVEILLANCE SR 3.7.12.2 REQUIREMENTS (continued) This SR verifies the Fuel Handling Area Ventilation System has not degraded and is operating as assumed in the safety analysis. The flow rate is periodically tested to verify proper function of the Fuel Handling Ventilation System.

When aligned to the "emergency filter bank", the Fuel Handling Area Ventilation System is designed to reduce the amount of unfiltered leakage from the fuel handling building which, in the event of a fuel handling accident, lowers the dose at the site boundary to well within the guidelines of 10 CFR 100. The Fuel Handling Area Ventilation System is designed to lower the dose to these levels at a flow rate of

~ 5840 cfm and ~ 8760 cfm. The Frequency of 18 months is consistent with the test for filter performance and other filtration SRs.

REFERENCES 1. FSAR, Section 9.8

2. FSAR, Section 14.11

3. FSAR, Section 14.19

4. Regulatory Guide 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurize Reactors.

5. NUREG-0800 Section 15.7.5, Spent Fuel Cask Drop

Accidents.

6. Regulatory Guide 1.52, De$ign, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorbtion Units of Light-Water-Cooled Nuclear Power Plants .

Palisades Nuclear Plant B 3.7.12-7 03/15/99

ESRV Dampers B 3.7.13 BASES LCO Two ESRV Damper trains are required to be OPERABLE to ensure that each engineered safeguards room isolates upon receipt of its respective high radiation alarm. Total system failure could result in the atmospheric release from the engineered safeguards rooms exceeding the required limits in the event of a Design Basis Accident (OBA).

An ESRV Damper train is considered OPERABLE when its associated radiation monitor, instrumentation, ductwork, valves, and dampers are OPERABLE.

APPLICABILITY In MODES 1, 2, 3, and 4, the ESR-Damper trains are required to be OPERABLE consistent with the OPERABILITY requirements of the Emergency Core Cooling System (ECCS).

In MODES 5 and 6, the ESRV Damper trains are not required to be OPERABLE, since the ECCS is not required to be OPERABLE *

ACTIONS Condition A addresses the failure of one or both ESRV Damper trains. Operation may continue as long as action is irrmediately initiated to isolate the affected engineered safeguards room. With the inlet and exhaust dampers closed, or if the inlet and outlet ventilation plenums are adequately sealed, the engineered safeguards room is isolated and the intended safety function is achieved, since the potential pathway for radioactivity to escape to the environment from the engineered safeguards room has been minimized.

The Completion Time for this Required Action is commensurate with the importance of maintaining the engineered safeguards room atmosphere isolated from the outside environment when the ECCS pumps are circulating primary coolant after an accident.

Palisades Nuclear Plant B 3.7.13-2 03/15/99

SFP Level B 3.7.14

B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool (SFP) Water Level BASES BACKGROUND The minimum water level in the SFP meets the assumptions of iodine decontamination factors following a fuel handling or cask drop accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the SFP design is given in the FSAR, Section 9.11 (Ref. 1), and the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.4 (Ref. 2). The assumptions of fuel handling and fuel cask drop accidents are given in the FSAR, Section 14.19 and 14.11 (Refs. 3 and 4), respectively.

APPLICABLE The minimum water level in the SFP meets the assumptions SAFETY ANALYSES of fuel handling or fuel cask drop accident analyses described in References 3 and 4 and are consistent with the assumptions of Regulatory Guide 1.25 (Ref. 5). The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to a person at the exclusion area boundary is well within the 10 CFR 100 (Ref. 6) limits.

Reference 5 assumes there is 23 ft of water between the top of the damaged fuel assembly and the fuel pool surface for a fuel handling or fuel cask drop accident. This LCD preserves this ass-umpti on for the bulk of the fuel in the storage racks. In the case of a single assembly, dropped and lying horizontally on top of the spent fuel racks, there may be < 23 ft of water above the top of the assembly and the surface, by the width of the assembly. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rods fail from a hypothetica.l maximum drop.

The SFP water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2) .

Palisades Nuclear Plant B 3.7.14-1 03/15/99

SFP Boron Concentration B 3.7.15

BASES APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool until a complete spent fuel pool verification of the stored assemblies has been performed following the last movement of fuel assemblies in the spent fuel pool. This LCO does not apply following the verification since the verification would confirm that there are no misloaded fuel assemblies. With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.

ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE l, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown .

A.1. A.2.1. and A.2.2 When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress. This is most efficiently achieved by i1T1T1ediately suspending the movement of fuel assemblies. This does not preclude the movement of fuel assemblies to a safe position. In addition, action must be inmediately initiated to restore boron concentration to within limit. Alternately, beginning a verification of the SFP fuel locations to ensure proper locations of the fuel can be performed.

Jalisades Nuclear Plant B 3.7.15-2 03/15/99

Spent Fuel Assembly Storage B 3.7.16

B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage BASES BACKGROUND The spent fuel storage facility is designed to store either new (nonirradiated) nuclear fuel assemblies, or used (irradiated) fuel assemblies in a vertical configuration underwater. The storage pool is sized to store 892 irradiated fuel assemblies, which includes storage for failed fuel canisters. The spent fuel storage racks are grouped into two regions, Region I and Region II per Figure 3.7.16-1. The racks are designed as a Seismic Category I structure able to withstand seismic events.

Region I contains racks in the spent fuel pool having a 10.25 inch center-to-center spacing and a single rack in the north tilt pit having a 11.25 inch by 10.69 inch center-to-center spacing. Region II contains racks in both the spent fuel pool and the north tilt pit having a 9.17 inch center-to-center spacing. Because of the smaller spacing and poison concentration, Region II racks have more limitations for fuel storage than Region I racks. Further information on these limitations can be found in Section 4.0, "Design Features." These limitations (e.g., enrichment, burnup) are sufficient to maintain a kett of ~ 0.95 for spent fuel of original enrichment of up to 4.40%.

APPLICABLE The spent fuel storage facility is designed for SAFETY ANALYSES noncriticality by use of adequate spacing, and "flux trap" construction whereby the fuel assemblies are inserted into neutron absorbing stainless steel cans.

The spent fuel assembly storage satisfies Criterion 2 of 10 CFR 50.36(c)(2).

LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Table 3.7.16-1, in the accompanying LCO, ensures that the keff of the spent fuel pool will always remain < 0.95 assuming the pool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Table 3.7.16-1, in the accompanying LCO. Euel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Specification 4.3.1.1 .

Palisades Nuclear Plant B 3.7.16-1 03/15/99

Spent Fuel Assembly Storage B 3.7.16

BASES APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II _of either the spent fuel pool or the north tilt pit.

ACTIONS Required Action A.l is modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

When the configuration of fuel assemblies stored in Region II the spent fuel pool is not in accordance with Table 3.7.16-1, inmediate action must be taken to make the

SURVEILLANCE necessary fuel assembly movement(s) to bring the configuration into compliance with Table 3.7.16-1.

SR 3.7.16.1 REQUIREMENTS This SR verifies by administrative means that the initial enrichment and burnup of the fuel assembly is in accordance with- Table 3.7.16-1 in the accompanying LCO prior to placing the fuel assembly in a Region II storage location.

REFERENCES None Palisades Nuclear Plant B 3.7.16-2 03/15/99

Secondary Specific Activity B 3.7.17

B 3.7 PLANT .SYSTEMS B 3.7.17 Secondary Specific Activity BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Primary Coolant System (PCS). Und~r steady state conditions, the activity is primarily iodines with relatively short half lives, and thus is indication of current conditions. During transients, I-131 spikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

This limit is lower than the activity value that might be expected from a 1 gpm tube leak of primary coolant at the

limit of 1.0 µCi/gm as assumed in the safety analyses with exception of the control rod ejection analysis which assumes 0.6 gpm. LCO 3.4.13, "PCS Operational LEAKAGE," is more restrictive in that the limit for a primary to secondary tube leak is 0.3 gpm. The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and primary coolant LEAKAGE. Most of the iodine isotopes have short half lives (i.e.,< 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ).

Operating a plant at the allowable limits would result in a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Exclusion Area Boundary (EAB) exposure well within the 10 CFR 100 (Ref. 1) limits .

Palisades Nuclear Plant B 3.7.17-1 03/15/99

Secondary Specific Activity

B 3.7.17 BASES APPLICABILITY In MODES 1, 2, 3, and 4, the limits on secondary specific activity qpply due to the potential for secondary steam releases to the atmosphere.

In MODES 5 and 6, the steam generators are not being used for heat removal. Both the PCS and steam generators are at low pressure or depressurized, and primary to secondary LEAKAGE is minimal. Therefore, monitoring of secondary specific activity is not requfred.

ACTIONS A.1 and A.2 DOSE EQUIVALENT I-131 exceeding the allowable value in the secondary coolant is an indication of a problem in the PCS and contributes to increased post accident doses. If secondary specific activity cannot be restored to within limits in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in 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. 7.17 .1 REQUIREMENTS This SR ensures that the secondary specific activity is within the limits of the accident analysis. A ganma isotope analysis of the secondary coolant, which de_termines DOSE EQUIVALENT I-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual i~otopic concentrations that might indicate changes in primary coolant activity or LEAKAGE. The 31 day Frequency is based on the detection of increasing trends of the level of DOSE EQUIVALENT I-131, and allows for appropriate action to be taken to maintain levels below the LCO limit.

Palisades Nuclear Plant B 3.7.17-3. 03/15/99

c. The fi water maKeuo o the Auxiliar. Feed~ater p, o Suc~ion (P-8 and P-88) may e inooeraole f a period of oays provia ~

th oumo service* er maKeup to P SC, oumo P-8C and its responding fl control valve are operable fire C.OIJD e, e*

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~E: 2 of= 4 Revised 03/15/99

m~y bt modified to low*prtssurt s1ftt 1 pu=p is rtstortd o proyided Yllvts, inttrloc or pipin9 direc y 1ssoci1ttd wit one of t 1bovt coac>ontn ind required to unction durin9 1 ident conditions sn1l dee11ed to~ p t of th1t c0111pon t ind shill

..,t tht S&81 1quirtmtnts 1s 11 td for th1t COlllP ent .

Any YllYt Interlock or pipt ssoci1ttd with t s1ftty inject* n ind sn own coolin9 syst ovtrtd under 3 .Zt abov ut, which is rtq td to function 1n9 accident it1ons, ~IY ~ in trablt for a per ~ of no more t n nours .

3*Z9a Amtndmtn*t No. ~. 9-t, 172 September Z6, 1996 Revised 03/15/99 I

The Control Room Ventilation and Isolation System {and/the j\jel~to/aga

!Area )!EP.A/CharJ?oa 1/Exhk!1st25tsfem I sha 11 be demonstrated to be OPERABLE by the following tests: ( ~e-0

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ontinued During rea.tor vessel head removal and while refueling operation are being performed in the reactor, the re boron c centration shall be maintained in the prima system nd shall be checked by sampling on each shi

h. Dire t co111T1unication between personnel in the con rel room and at he refueling machine shall be available whe ver changes in c e geometry are taking place.

3.8.2 If a y of the conditions in 3.8.1 are not met, refueling ope ations shall cease i111T1ediately, work shall b initiated to s isfy the required conditions and no operatic s that may change e reactivity of the core shall be made.

3.8.3. Refueling operation shall not be initiated fore the reactor core has decayed for a minimum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if t e reactor has been operated at power levels in excess of 2% ated wer. ~

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RA A.l 3.8.5. Wen spen ue w ic as ecaye ess an one year s p aced in the tilt it storage racks, the bulk water temperat e in the tilt pit star ge area must be monitored continuously to ssure that the water t mperature does not exceed 1S0°F. Monitori g will continue for 24 ours after any addition of fuel to them n pool or the tilt pit o when a failure of the spent fuel pool coo ing system occurs.

ipment and general procedures to be utili d during refueling are disc sed in the FSAR. Detailed instructions, e above specifications, and he design of the fuel handling equipment i corporating built-in in rlocks and safety features provide assura e that no incident could o ur during the refueling operations that w ld result in a hazard to blic health and safety.(l) Whenever chan s are not being made in core eometry, one flux monitor is sufficient. his

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3-47 Amendment No. ~~, 81 Revised 03/15/99

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ng limit1tions dur g refueling oper Qbjtctivt 3.8.1 jse:.E *I

\~*~ I LCO 5.1.IZ.

b. All ut09lt1c contain t 1sol1t1on valvt shall be optrable rat 1 st ont valve in 1 11nt shall bt cl Id *

nt1ng and PU1"91 SY. ..s, including t radiation tiat1 isolation, all be ttstld and tr1fitd to both bt op1r t 111111d1attly pr1 to rtfutling opt tions. The two monitor. shall bt located th1 conta1n..nt f tl handling area 11vtl (tl ation 6'9'), shal be part of th* pl arta monitoring syst.. d shall ...,loy on out-of-two logic f isolation. Duri nonn optrat1on, thtsl nitors will not in htt an isohtion

1. A switch shal t provided so tha isolation action n be tiatld during rtf ing only.

Radiation 11v1ls n th1 contain..nt shall be 11anit Id continuously.

hanged, neutron f x shall be cont1nuo 1 11anitorld by at ast two sourct ra 1 neutron monitor., with 11ch 110nitor rov1d1ng cont1nuo visual indicat* n in t control roOll. Wh1 core geometry is t being changed at 11 one source range utron 11anitor sha be in s1rv1c1.

t l11st one shutd in op1r1t1on *

3-46 Amendment No. 34 January 27. ; 373 Revised 03/15/99

src:c1 F:c.A"TiorJ ~ .1. 1(.

4.2 VENTI ~TION SYSTEM TESTS S.R :i.=t. I 2.. I At least once per refueling cycle by:

\

a. Verify* g that on a containment high-pres ure and high-radia on test signal. the Control Room entilation system auto tically switches into the emergen y mode of operation wit flow through the HEPA filter and arcoal adsorber ba .

\ b. V rifying that the Control Room Ven ilation system maintains he Control Room at a positive pre sure ~ 1/8 inch WG re 1at i ve to the outside atmospher dur.i ng system emergency mode o eration.

c. Verifying that the Fuel Pool Ventilation System is OPERABLE by initiating flow through the HEPA filter and charcoal adsorbers from the control room.

< A'D~ SR 3.1-. 12..'2.)

4-14 Amendment No. 8+/-. ~. -+/--74. Revised 03/15/99

A.6 SPECIFICATION 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM ATTACHMENT 3 DISCUSSION OF CHANGES A Note has been added to SR 3.7.5.3 which states "Not required to be met in MODES 2 or 3 when AFW is in operation." This Note is needed to prevent unnecessary entering of ACTIONS for LCO 3.7.5 during the startup or shutdown of the plant for not being able to meet the SR. Palisades uses the AFW system for steam generator level control during startup and shutdown in MODES 2, 3, and 4. During these operations the flow control valves used are in manual, and will not open automatically when an actuation signal is received, which would fail the SR. This change is administrative because CTS 4.9.b.1 states "each valve to actuates to its correct position (or that the specified flow is established) upon receipt of a simulated auxiliary feedwater pump start signal." During startup or shutdown the valves are providing the proper flow for the existing plant condition. This Note is appropriate because the valves are needed to be throttled in these conditions to prevent overfilling of the steam generators due to low steam flow conditions, also the Note clarifies current licensing basis requirements.

A.7 This change adds the additional "inservice requirements" as described in ASME Code,Section XI to CTS 4.9.a.1 and 2. This change is administrative in that these requirements are performed by current surveillances and also this change only combines the two requirements, Code and TS. This change is consistent with NUREG-1432.

A.8 CTS 3.5.4 provides corrective actions in the event all AFW pumps are inoperable. In this case, the capability to provide the required AFW flow lo either steam generator has been lost. Proposed ITS 3.7.5 Condition D also provides corrective actions when the capability to provide the required AFW flow to either steam generator has been lost.

Condition D is stated as "two AFW trains inoperable with both steam generators having less than 100 % of the AFW flow equivalent to a single Operable AFW train available in Mode 1, 2, or 3 or, (the) required AFW train inoperable in Mode 4." Since the AFW system inoperabiliy addressed in ITS 3.7.5 Condition D (a loss of AFW function) is equivalent to the condition presented in CTS 3.5.4, this change has been characterized as administrative in nature .

Palisades Nuclear Plant Page 2 of 7 03/15/99

ATTACH1\1ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM ADMINISTRATIVE CHANGES (A)

A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.

Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.

- A.2 CTS 3.4.2 and 3.4.3 require that if a component(s) listed in Specification 3.4.1 is inoperable for more than the time specified, the plant must be placed in HOT SHUTDOWN. In proposed ITS 3.7.7 Required Action B.1, the CTS term is replaced with MODE 3. This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

A.3 CTS 3.4.4 specifies that valves, interlocks and piping that are directly associated with the "above" (CTS 3.4.1) components shall meet the same requirements as listed for that component. CTS 3.4.5 specifies that valves, interlocks and piping which is associated with the containment cooling system and not covered by CTS 3.4.4 may be inoperable for no more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if it is required to function during an accident. These requirements are addressed by the definition of OPERABILITY which requires that all associated equipment be OPERABLE. In the proposed ITS, all equipment in a particular train which is required to function during an accident must be OPERABLE and all equipment in the train will have the same Completion Time. This is an administrative change since the requirement remains that all equipment in a train of containment cooling must be OPERABLE. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 1of5 03/15/99

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.7, COMPONENT COOLING WATER (CCW) SYSTEM A.4 CTS 3.3.2, 3.4.2, and 3.4.3 require that with the Required Action and associated Completion Time not met the plant must be placed in COLD SHUTDOWN. In proposed ITS 3.7.7 Required Action B.2, the CTS term is replaced with MODE 5.

This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

A.5 CTS 3.4.3 states " .... Continued power operation with one component out of service J

shall be as specified in Section 3.4.2, with the permissible period in inoperability starting at the time that the first of the two components became inoperable." This explanatory information on the usage rules of technical specifications is addressed in the proposed ITS Section 1.3, "Completion Times," and does not need to be addressed in the Actions of proposed ITS 3. 7. 7. This is considered to be an administrative

I change since the requirements on complying with the completion times is addressed in the proposed ITS. This change is consistent with NUREG-1432.

A. 6 The Note added to proposed SR 3. 7. 7. 1 to aid the operator in the prevention of entering an inappropriate LCO. The Note reminds the operator that loss of CCW flow to a component may render that component inoperable but does not affect the OPERABILITY of the CCW System. This change is considered administrative that this is a clarifier to the operator to prevent confusion. This change is consistent with NUREG-1432.

TECHNICAL CHANGES

MORE RESTRICTIVE (M)

M.l CTS 3.3.1, 3.3.2, 3.4.1, and 3.4.2 establish the Applicability for' the various components which comprise the CCW by stating that "the reactor shall not be made critical. ... unless all of the following conditions are met." The Applicability of the CCW in proposed ITS 3.7.7 is MODES 1, 2, 3, and 4. As such, the requirements associated with CTS 3.3.1, 3.3.2, 3.4.1, and 3.4.2 have been revised to be more restrictive by requiring the CCW to also be OPERABLE during the additional MODES 3 and 4. SRs 3.7.7.2 and 3.7.7.3 are modified by a Note which states that these SRs are not required to be met in MODE 4. This is due to the instrumentation providing the signals are not required in MODE 4. '!_'his change keeps consistency with ITS 3. 3. 3, "ESF Instrumentation," and current licensing basis. This change is an additional restriction on plant operations and is consistent with NUREG-1432.

Palisades Nuclear Plant Page 2 of 5 03/15/99

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.10, CRV FILTRATION LESS RESTRICTIVE CHANGES (L)

L.1 CTS 4.2, Table 4.2.3, item 2.a requires a verification that the Control Room Ventilation system automatically switches into the emergency mode of operation on a "containment high pressure and high radiation test signal." The Applicability of this requirement is "above COLD SHUTDOWN, during REFUELING OPERATIONS, during movement of irradiated fuel assemblies, and during movement of a fuel cask in or over the Spent Fuel Pool." Proposed SR 3.7.10.3 requires a verification that each CRV Filtration train actuates on an actual or simulated actuation signal. The requirement and Applicability of CTS 4.2, Table 4.2.3, item 2.a is similar to the requirement and Applicability of SR 3. 7 .10. 3. However, SR 3. 7 .10. 3 is further modified by a Note which states that the SR is "not required to be met during movement of irradiate fuel assemblies in the SFP, or during movement of a fuel cask in or over the SFP. " The purpose of this Note is to exclude the requirement of the SR during those plant evolutions in which no instrumentation is available to actuate the CRV System. The CRV System is designed to automatically switch to the emergency mode of operation on a "containment high pressure or containment high radiation signal." The instruments used to initiate these actuation signals are not capable of

detecting an increase in radiation levels in the fuel handling building, and as such, can not provide automatic actuation of the CRV System in the event of a fuel handling accident or cask drop accident in the SFP. Therefore, the addition of the Note in SR 3.7.10.3 establishes consistency with the design of the CRV System and the requirement of the SR. During movement of irradiate fuel assemblies in the SFP, or during movement of a fuel cask in or over the SFP, manual operator action is necessary to initiate the emergency filtration mode of the CRV System .

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ATTACHMENT 3 DISCUSSION OF CHA.t"lGES SPECIFICATION J. 7.12, FUEL HANDLING AREA VENTILATION SYSTEM ADMINISTRATIVE CHANGES (A)

A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.

Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.

A.2 CTS 3.8.4 requires that the "ventilation system and charcoal filter" be in operation during specified conditions. Proposed LCO 3. 7 .12 has the same requirements, but replaces "charcoal filter ... be operating" with the phrase "aligned in the emergency filtration mode with one exhaust fan in operation." This change is considered editorial in that the wording describes the same components and provides a clearer description to the operator of what needs to be performed.

A.3 The CTS definition of REFUELING OPERATION forms the basis for the proposed ITS definition of CORE ALTERATIONS. This change is considered to be administrative since the term "CORE ALTERATIONS" is used to simply replace "REFUELING OPERATION," any additional clarification provided by the new definition gives additional clarification on its application. This change is consistent with NUREG-1432.

A.4 CTS 3.8.4 requires fuel movements be terminated if both (Fuel Handling Area exhaust) fans are unavailable "until one fan is returned to service." Proposed ITS 3.7.12 also requires fuel movements be suspended if the Fuel Handling Area Ventilation system is not in operation (i.e., one exhaust fan running) but, does not contain the stipulation "until one fan is returned to service." Inclusion of the phrase "until one fan is returned to service" is unnecessary since once a fan is returned to service compliance with the LCO is restored. Thus, omission of this phrase is considered administrative in nature since it does not alter the original intent of the CTS requirement. This change is consistent with NUREG-1432.

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ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.12, FUEL HANDLING AREA VENTILATION SYSTEM TECHNICAL CHANGES - MORE RESTRICTIVE (M)

I M. l The requirements of CTS 3.8.4 have been revised to address the Operability of the Fuel Handling Area Ventilation system. LCO 3.7.12 requires, in part, that the Fuel Handling Area Ventilation system must be Operable. Although CTS 4.2, Table 4.2.3 item 2.c contains a test to verify the Operability of the Fuel Pool Ventilation System, the CTS does not contain an explicit requirement for Operability. As such, a new Condition, Required Action and Completion Time has been provided to address the situation when the Fuel Handling Area Ventilation system is inoperable. This change is consistent with NUREG-1432.

M.2 CTS 3.8.4 requires Fuel Handling Area Ventilation System to be in operation during REFUELING OPERATIONS with the equipment door open and also fuel handling in the fuel storage building, if fuel in either locations has decayed < 30 days. Proposed ITS 3. 7.12 requires Fuel Handling Area Ventilation System in operation during CORE _

ALTERNATIONS (see DOC A.3) with the equipment door open, movement of irradiated fuel assemblies in the containment with the equipment door open, and

movement of irradiated fuel assemblies in the fuel handling building, if fuel in either location has decayed < 30 days. In addition, proposed ITS 3.7.12 requires the operation of the Fuel Handling Area System during movement of a fuel cask in or over the spent fuel pool when irradiated fuel assemblies with < 90 days decay time are in the fuel handling building. This change clarifies the conditions when Fuel Handling Area Ventilation System shall be in operation and adds the new requirements for Fuel Handling Area System operation. The Design Basis Accidents are fuel handling accidents and a cask drop accident, both of which involve the release of airborne radioactive particles to the fuel handling area. The change is considered more .

restrictive because of the added requirement on plant operation.

M.3 CTS does not have a specific Surveillance for Fuel Handling Area Ventilation System as presented in proposed ITS SR 3.7.12.3. The proposed surveillance requires additional testing to be performed on the Fuel Handling Area Ventilation System to verify OPERABILITY. The proposed SR verifies that the Fuel Handling Area Ventilation System has not degraded and is operating at the flow rate assumed in the analysis. This change is considered more restrictive since the system is.not required by CTS to be tested in this manner .

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SPECIFICATION 3.7.12, FUEL HANDLING AREA VENTILATION SYSTEM LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

ATTACHMENT 3 DISCUSSION OF CHANGES LA.1 CTS 3 .8.4 states if both (Fuel Handling Area Ventilation exhaust) fans are unavailable, then "any fuel movements in progress shall be completed ... ". The intent of this statement is to clarify that the actions do not preclude the movement of a fuel assembly to a safe position which ultimately minimizes the potential for a fuel handling accident.

In proposed ITS 3.7.12 this same clarification is provided in the Bases. Placing this information in the Bases is acceptable since these details are not pertinent to the actual requirements, but merely describe safe operating practices. 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 4.2, Table 4.2.3 item 2.c requires a verification "that the Fuel Pool Ventilation System is Operable by initiating flow through the HEPA filter and charcoal adsorbers from the control room at least once per refueling cycle." In proposed ITS 3. 7 .12, this surveillance requirement has been deleted since it is redundant to the actual requirement of the LCO. LCO 3. 7 .12 requires that the Fuel Handling Area Ventilation System be OPERABLE and aligned in the emergency filtration mode with one exhaust fan in operation. As such, in order to establish compliance with the LCO, flow must be initiated through the emergency filter bank which includes the HEPA and charcoal adsorbers. Specifying that flow be initiated from the control room is irrespective of the safety function performed by the Fuel Handling Area Ventilation System since the system must be aligned in the emergency filtration mode prior to movement of any irradiated fuel assemblies. Therefore, since the requirement of LCO 3. 7 .12 fulfills the requirement CTS 4.2, Table 4.2.3 item 2.c, this surveillance requirement can be deleted without an impact of safety .

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ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY ADMINISTRATIVE CHANGES (A)

A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.

Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.

A.2 The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content consistent with NUREG-1432. The revised Bases are shown in the proposed Technical Specification Bases.

A.3 CTS 3.l.5c requires that with specific activity of the secondary coolant >0.1 µCi/gram DOSE EQUIVALENT I-131, the plant must be placed in HOT SHUTDOWN. In proposed ITS 3.7.17 Required Action A.l, the CTS term is replaced with MODE 3.

This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

A.4 CTS 3. l.5c requires that with specific activity of the secondary coolant > 0.1 µCi/gram DOSE EQUIVALENT 1-131, the plant must be placed in COLD SHUTDOWN. In proposed ITS 3.7.17 Required Action A.2, the CTS term is replaced with MODE 5.

This is considered to be an administrative change since the effect on operations is similar. This change is consistent with NUREG-1432.

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ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY A.5 CTS 3.1.5c requires that with specific activity of the secondary coolant >0.1 µCi/gram DOSE EQUIVALENT 1-131, the plant must be placed in COLD SHUTDOWN. In proposed ITS the term is replaced with MODE 5 (see DOC A.4). In proposed ITS 3. 7 .17 Applicability, the Specification is applicable in MODES 1, 2, 3. and 4.

Placing the plant.in COLD SHUTDOWN in CTS and having the Applicability in MODES 1, 2, 3, and 4 in proposed ITS is basically the same. This change is considered to be an administrative change since the effect on operations is similar.

This change is consistent with NUREG-1432.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M.1 CTS 4.2 Table 4.2.1, item 7a, requires the specific activity of the secondary coolant system to be determined once per 31 days when~ver the gross activity determination indicates iodine concentrations greater than 103 of the allowable limit, and once per

6. months whenever the gross activity determination indicates iodine concentrations below 10 3 of the allowable limit. Proposed ITS SR 3. 7 .17 .1 will require the specific activity to be determined once per 31 days. The proposed ITS SR will not contain the allowance to extend the SR interval to 6 months whenever the gross activity determination indicates iodine concentration below 103 of the allowable limit. This change does not adyersely affect safety because the 31 day interval ensures that the specific activity is checked frequently enough to establish a trend to identify secondary activity problems in a timely manner. Deleting an allowance to extend an SR interval constitutes a more restrictive change. This change is consistent with NUREG-1432.

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

There were no "Removal of Details" associated with this specification.

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ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY LESS RESTRICTIVE CHANGES (L)

L.1 CTS 4.2, Table 4.2.1 requires a sample of secondary coolant to be analyzed for gross radioactivity 3 times every 7 days with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples. The CTS contains no LCO, limiting value, or Required Actions for secondary coolant gross radioactivity, only that sampling is required. The intent of this surveillance is to monitor the iodine concentration in the secondary coolant in order to determine the frequency at which an isotopic analysis for Dose Equivalent I-131 concentration in the secondary coolant is performed. The CTS requires an isotopic analysis for Dose equivalent I-131 of the secondary coolant once per 31 days whenever the gross activity indicates iodine concentrations greater than 10% of the allowable limit or, once per 6 months whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. However as discussed in DOC M.1 for this specification, the extended surveillance interval of 6 months for the determination of Dose Equivalent I-131 in the secondary coolant has been proposed for deletion and that future testing be performed every 31 days. Thus, the need to perform sampling of the secondary coolant for gross radioactivity is no longer necessary and has been delete in the ITS. This change is acceptable since gross radioactivity in the secondary coolant is not evaluated for radiological consequences in any of the accidents assumed in the FSAR, and the concentration of the Dose Equivalent I-131 in the secondary coolant will continue to be determined at an appropriate frequency. In addition, radiation monitoring instrumentation, controlled in accordance with the Offsite Dose Calculation Manual (e.g., SG blowdown monitors and condenser off gas monitor), is available to monitor increases in the radioactivity levels in the secondary coolant. This change is consistent with NUREG-1432 .

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_J

NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.7.12, FUEL HANDLING AREA VENTILATION SYSTEM LESS RESTRICTIVE CHANGE L.1 ATTACHMENT 4 CTS 4.2, Table 4.2.3 item 2.c requires a verification "that the Fuel Pool Ventilation System is Operable by initiating flow through the HEPA filter and charcoal adsorbers from the control

- room at least once per refueling cycle." In proposed ITS 3. 7 .12, this surveillance requirement has been deleted since it is redundant to the actual requirement of the LCO. LCO 3.7.12 requires that the Fuel Handling Area Ventilation System be OPERABLE and aligned in the emergency filtration mode with one exhaust fan in operation. As such, in order to establish compliance with the LCO, flow must be initiated through the emergency filter bank which includes the HEP A and charcoal adsorbers. Specifying that flow be initiated from the control room is irrespective of the safety function performed by the Fuel Handling Area Ventilation System since the system must be aligned in the emergency filtration mode prior to movement of any irradiated fuel assemblies. Therefore, since the requirement of LCO 3. 7 .12 fulfills the requirement CTS 4.2, Table 4.2.3 item 2.c, this surveillance requirement can be deleted without an impact of safety.

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 eliminates the requirement to perform a test on the Fuel Pool Ventilation System on the basis that the intent of the test is adequately fulfilled by complying with the associated LCO prior to establishing a condition where the system would be required to function. The proposed change does not involve a change to any accident initiators or precursor. Therefore, this change 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 function or operability of the Fuel Pool Ventilation System. As such, the consequences of an accident involving operation of the Fuel Pool Ventilation System remain unchanged. Therefore this change does not involve a significant increase in the consequence of an accident previously evaluated.

Palisades Nuclear Plant - Page 1of2 03/15/99

2.

SPECIFICATION 3.7.12, FUEL HANDLING AREA VENTILATION SYSTEM ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION 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. The proposed change does not alter the accident mitigative function of the Fuel Pool Ventilation System, nor does it create an opportunity for new or different accident beyond those previously evaluated. Therefore, 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 equipment design, operating parameters, and the set points at which automatic actions are initiated within analyzed limits. There are no design changes or equipment performance parameter changes associated with this change. No accident or transient analysis are affected by this change. Therefore, this

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

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY LESS RESTRICTIVE CHANGE L.1 CTS 4.2, Table 4.2.1 requires a sample of secondary coolant to be analyzed for gross radioactivity 3 times every 7 days with a maximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples. The CTS contains no LCO, limiting value, or Required Actions for secondary coolant gross radioactivity, only that sampling is required. The intent of this surveillance is to monitor the iodine concentration in the secondary coolant in order. to determine the frequency at which an isotopic analysis for Dose Equivalent I-131 concentration in the secondary coolant is performed. The CTS requires an isotopic analysis for Dose equivalent I-131 of the secondary coolant once per 31 days whenever the gross activity indicates iodine concentrations greater than 10 % of the allowable limit or, once per 6 months whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. However as discussed in DOC M .1 for this specification, the extended surveillance interval of 6 months for the determination of Dose Equivalent I-131 in the secondary coolant has been proposed for deletion and that future testing be performed every 31 days. Thus, the need to perform sampling of the secondary coolant for gross radioactivity is no longer necessary and has been delete in the ITS. This change is acceptable since gross radioactivity in the secondary coolant is not evaluated for radiological consequences in any of the accidents assumed in the FSAR, and the concentration of the Dose Equivalent I-131 in the secondary coolant will continue to be determined at an appropriate frequency. In addition, radiation monitoring instrumentation, controlled in accordance with the Offsite Dose Calculation Manual (e.g., SG blowdown monitors and condenser off gas monitor), is available to monitor increases in the radioactivity levels in the secondary coolant. 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 deletes the sample requirement for gross radioactivity of the secondary coolant. This sample does not have a detrimental impact on the integrity of any plant structure, system, or component. Deletion of this sample requirement will not alter the operation of any plant equipment, or otherwise increase its failure probability. As such, the probability of occurrence for a previously analyzed accident is not significantly increased .

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.7.17, SECONDARY SPECIFIC ACTIVITY 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. Gross radioactivity of the secondary coolant is not an initial condition input assumed for any analyzed event.

The amount of Dose Equivalent I-131 in the secondary coolant is the assumed parameter. The limit requirement for Dose Equivalent I-131 remains unchanged and the sampling requirement has become more restrictive (see M. l). The deletion of the gross radioactivity sampling requirement does not affect the assumptions of an analyzed event. This change does not affect the perlormance of any credited equipment since the sample requirement is for an unassumed parameter. 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.

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 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 set points 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. Deleting the sample requirement for gross radioactivity of the secondary coolant does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, ~o 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 set points at which automatic actions are initiated. Deleting the requirement to sample the secondary coolant for gross radioactivity does not significantly impact these factors. There are no design changes or equipment performance parameter changes associated with this change. Therefore, this change does not involve a significant reduction in the margin of safety.

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UHS 3.7.9

3.7 PLANT SYSTEMS 3.7.9 Ultimite Heat SinK (UHS)

LCO 3.7.9 The UHS sha 11 be OPERABLE.

APPLICABILITY: MOeES l, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION r'IME One or re cooling A. l Res re cooling tower 7 days L towers ith one coolin tower fan inope able.

A ~

fan s) to OPERABLE st tus.

~l Be in MOOE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

~

@.2 Be in MOOE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months A. ~

Q)

SURVEILLAMCI A( IREMENTS SURVEILLANCE FREQUENCY

~ Verify water 1evel of UHS is ?: (ffiI! ft ~ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

. 3.7.9.l

~ean sei level)':' _t (continued)

CEOG STS 3.7-21 Rev 1, 04/07/95 Revised 03/15/99

tfiG$ PREACS\

3. 7. l 3 LCO 3.7.13 Two ECCS PREACS trains shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

fi riio~

On ECCS PREACS trair\S A. l [4avi !~lQ~Y rJ e.v.J inoperible.

<0 ~"'* +ro.; ".

8. Req red Action and Be in MOOE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months iSS ciited Completion Ti not met.

Be in MOOE SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Operate each ECCS PRE S train for 31 days

[~ 10 continuous ho s with the heater operating or (for stems without heate

~ 15 minutes].

(continued)

CEOG STS 3.7-29 Rev 1, 04/07 15 Revised 03/15/99

FBA~ @

3. 7.ql) 12.

~ LCO 3.7.(1~ (two FBAC¥ trains sh,11 be OPERAQlE .)

~~I ,7[::~

INS~, '- ~

~ APPLICABILITY: [MODES , 2, 3, and 4,]

Durin movement of irrad ated fuel assemblie in the fuel

.___,_u"'-'n ding.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.15 A. One FBACS train A.I Restore FBACS train 3.15.L.{ ___ inoperable. to OPERABLE status.

!N.l<RT

~ 4.1 I B. Required Ac:ti B.l Be in MODE 3.

e and assoc:iat Co~letion T G) I of Conditio A not met in MOOE 1, 2, 3, or 4.

Mil B.2 Be in MODE 5.

Two FBAC trains inopera e in MOOE 1, 2, 3, 0 4.

c. Requ Action and C. l Place OPERABL FBACS Inmediately Ass 1ated COlll)let1on train in oper tion.

T1 [of Condition A]

no *t during QB

...nt of irradiated f *1 asslllb11es in the C.2 Suspend ement of . Inned i ate 1y

1 building. irradiate fuel assellbli s in the fuel bu ding.

(continued)

@G Sl9 3.7-31 Rev 1, 04/07/95

'?ahsc:a.J4.s Revised 03/15/99

SECTION 3.7 INSERT I The Fuel Handling Area Ventilation System shall be OPERABLE with one fuel handling area exhaust fan, aligned to the emergency filter bank. in operation.

INSERT 2 During movement of irradiated fuel assemblies in the fuel handling building when irradiated fuel assemblies with < 30 days decay time are in the fuel handling building, During movement of a fuel cask in or over the SFP when irradiated fuel assemblies with < 90 days decay time are in the fuel handling building, During CORE AL TERATIO NS when irradiated fuel assemblies with< 30 days decay time are in the containment with the equipment hatch open, During movement of irradiated fuel assemblies in the containment when irradiated fuel assemblies with< 30 days decay time are in the containment with the equipment hatch open.

INSERT 3 CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel Handling Area A.1 Suspend movement of Immediately Ventilation System fuel assemblies.

not aligned or in operation. A.ND QR A.2 Suspend CORE ALTERATIONS. Immediately Fuel Handling Area Ventilation System A.ND inoperable.

A.3 Suspend movement of Immediately a fuel cask in or over the SFP .

3.7-31 Revised 03/15/99

FBACS 3.7.GIZ.

ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME D. ains D.l Suspend ovement of inoperabl during irradi ed fuel movemen of irradiated asse ies in the fuel a emblies in the fuel uil ding.

fuel ilding.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Operate each FBACS tr in for

[~ 10 continuous ho s -ith the heaters operating or (for s stems without heaters)

~ 15 minutes].

I

-7T~ ~.2.

SR 3.7.14.(/) Perfo,,. required-SAC filter testing in In accordance accordance with 1 ilation Filter with the~

-(61- LJ z. ' Testing Program VFXP * ~l.&..-h11 F1L+-c...t 0

k-1"' I ~s.4'1n r°'tlh,v"t" Verify each FBACS trai on an actual or simul signal.

SR 3.7.14J (continued)

CEOG STS 3.7-32 Rev l, 04/07/95

Revised 03/15/99

FBACS 3.7.@12.

SURVEILLANCE REOUIREHENTS continued SURVEILLANCE FREQUENCY E 3.7.14.S Verify each FBACS filter by ass diJ11Per can be opened .

CEOG STS 3.7-33 Rev l, 04/07/95

Revised 03/15/99

Spent Fuel Assembly Storaj.!.

3.7 3.7.~

PLANT SYSTEMS Spent Fuel Assembly Storage

~ 3.7.~

APP LI CAB IL ITV : Whenever any fuel assembly is stored in [Region~ oft\ the sP&.,t fuel storage poo~ . vf+..c.<<.

o<'-tht ~ +ii+,,+

ACTIONS ro ~

CONDITION

"' ~ "-.. REQUIRED ACTION COMPLETION TIME A. Requirements of the A. l l - - - - - - - -NOTE - -- - - -- - -

LCO not met. LCO 3.0.3 is not appl iCible.

Initiate iCtion to I111118diately move the noncomplying

(!) fuel assembly from

@ '(Region fll-t' 1t SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY crS '"'

3.7.~.l Verify by adll1ntstrat1ve means the initial enrtchMnt and burnup ~* fuel assembly Prior to storing the s.~:t.J 1~ 1n accordance with

3. 7 .Ga-1 IJ fuel assembl~

.5 ecjftcatilD 4.3.1/lj. "Tl, in ~egion tis 1t

<D

CEOG STS 3.7-39 Rev 1, 04/07 95 Revised 03/15/99

- B 3 . 7 Pl.AMT SY.STEMS B 3.7.2 Main Steam Isolation Valves (MSIVs)

MS IVs B 3.7.2 BASES 1-/ ~~STe~ ot: ~MS.IV BACKGROUND The MSIVs isolate steam flow f}'Oll t~e seco~darY. side of the steUI generators following a ~igh inenJy 1ine ireak (HELB .~

MSIV closure tenainates flow frOll the unaffected (intact) steUI generator~~ t3f'E:lir.~ ~~OF'* oT1o1o~ ""~il/J© One MSIV is located in each main steam line outside, but close to, containmeJ1t. The MSIVs are downstream from the

~ A1'11n iteam 'afety ialves (MSSVs), atmospheric dump valves, and auxiliary feedwater pwap turbine steaa supplies to prevent their beii.~ isolated frOll the steu generate.

MSIV closure. Closing the MSIVs isolates each steam TuRat~

@G1i.L.~) ~ypass generator from the other, and isolates the turbine,

$fftll. and other auxiliary steam supplies from the steUI generators~ r@

@~~er1>

The MSIVs~l se on l/11!iili it"' 1soht1on si by either ~steam gener1tor,pressur1 or g onta* nt na~enented@

GJ ressure. M V fail closed on loss of 1

a The also actuates the ain eedwa er~

i o alves (H~s)/jo close. The MSIVs m"'~~o be ac:tua ted manua 11 y. C.(g) ~ f.Jrf.v 5YP1.4>> v.6.t..\IB~

A description of the MSIVs is found in the FSAR,

[]J Stet1on_,fl0~(Ref. 1) *

CEOG STS B 3.7-7 Rev 1, 04/07/95 Revised 03/15/99

SECTION 3.7 INSERT 1

..... assuming the normally closed MSIV bypass valves are closed. The MSIV bypass valves do not receive an isolation signal and might be open during zero power conditions.

INSERT 2 The MSIVs are swing disc check valves. The inherent characteristic of this type of valve allows for reverse flow through the valve on a differential pressure even if the valve is closed.

In the event of an MSLB, if the MSIV associated with the unaffected steam generator fails to close, both steam generators may blowdown. This failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a double steam generator blowdown event, although more severe than the MSLB used in the original licensing basis of the plant, is not expected to result in unacceptable consequences.

Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial improvement to plant safety.

There are three different limiting MSLB cases that have been evaluated, one for fuel integrity and two for containment analysis (one for containment temperature and one for containment pressure). The limiting case for containment temperature is the hot full power MSLB inside containment following a turbine trip. At hot full power, the stored energy in the primary coolant is maximized .

B 3.7-7 Revised 03/15/99

MS IVs B 3.7.2 BASES APPLICABLE The accident analysis compares several different SLB eve s against different acceptance criteria. The ltir.ii SLB M outsid1 containment upstrea111 of the MSIY is limiting for 7 offsite dose, although a break in this short section of main steui header has a very low probability. The 1IJr11 SLB insidt containment at hot rzi1il power is the li*itin9 case for a post trip return to power. Tht analysis includes scenarios with offsite power available and with a loss of offsit1 power following turbine trip.

With offsit1 power available, the t.Ji,11-.1.1.1 continue to circulate coolant through The MSIYs serve only a safety function and r~in open during power operation. Thes1 valves op1rate under the

__....llowing situations:

MC)L6 (continued)

CEOG STS B 3;7-8 Rev 1, 04/07/95 Revised

03/15/99

SECTION 3.7 I.NSERT 1 With the most reactive control rod assumed stuck in the fully withdrawn position, there is an increased possibility that the core will return to power. The core is ultimately shut down by a combination of doppler feedback, steam generator dry out, and borated water injection delivered by the Emergency Core Cooling System .

B 3.7-8 Revised 03/15/99

SECTION 3.7 INSERT 1

..... and MFRV bypass valves in conjunction with feed pump speed, control Main Feedwater (MFW) flow to the steam generators for level control during normal plant operation. The valves also INSERT 2 The MFRVs and MFRV Bypass valves are non-safety grade valves located on non-safety grade piping that fails "as-is" on a loss of air. If required, MFW isolation can be accomplished using manually operated valves located upstream or downstream of the MFRVs and MFRV Bypass valves. In addition, each MFRV is equipped with a handwheel that can be used to isolate this MFW flowpath.

B 3.7-13 Revised 03/15/99

MFIVs [and [MFIVJ Bypass V1lves]

8 3. 7. :,

BASES

?Z:ie> M~ftN '8i(PA..~ v*1.ves)

BACKGROUND A description of the MFIVs'17 found in the FSAR, (continued) Section~ l O * (Ref. 1) *.__-----:---::-:-;:;;:~

?...3 APPLICABLE SAFETY ANALYSES l\ o Q ul'l\;J *I\ 4,, (. ~r-----4~~ """'7-F'"==":...;.:----=.::~_:_::rr;:..i:.=r--:.~~;..;;:..i

! ~ rt sPerlJC. c. Du8)

\ ::i.,....,., sir LCO on 15' TW'E:-R.

APPUCAl ILITY Tht MFIVs and bypass valves must bt OPERABLE whenever there is significant mass and energy in the~Coolant

@~

(continued)

CEOG STS B 3.7-14 Rev 1, 04/07/95 C.1..0SC.t:> A~D OE:~'T"l\/A'TE'C:>, Ole. ISOl..ATE.l:>

~L.O~~O MAi..1.LlALLV A,;:.T\),AiT~C> VAL~,

Revised

01/151_99

SECTION 3.7 INSERT However, this failure was not analyzed as part of the original licensing basis of the plant. As such, a Probabilistic Risk Assessment and cost benefit analysis were performed to determine if a facility modification was needed. The results of the analysis as described in an NRC Safety Evaluation dated February 28, 1986 concluded that a single steam generator blowdown event with continued feed water, although more severe than the MSLB used in the original licensing basis of the plant, is pot expected to result in unacceptable consequences. Furthermore, the NRC evaluation demonstrated that the potential offsite dose consequences are low and that modifications would not provide a cost beneficial iI1].provement to plant safety .

B 3.7-14 Revised 03/15/99

MFIVs [and [MFIV) Bypass Valves]

B 3.7.3 BASES APP LI CAB I LI TY (continued)

@ Q

-eiOT~ M.'t=rl...v At.Jt::I

~cmi- 1-.A.!=-~v 6'l'~!t VA1..v'&S A~:.

~1~E~ 1 ACTIONS (continued)

CEOG STS B 3.7-15 Rev 1, 04/07/95

Revised

03/15/99

MFIVs [and [MFIV] Bypass Valves]

B 3.7.3 BAs.ES rc..YMSc. ca."' C.c:rt.

rc.li>on~ bN6)

SR 3.7.3.1 (continued)

_ _....c-<io. t. ~J ~

REFERENCES 1. FSAR, Sect1on (c)OJ./i. '" ;,_:,..;

2. ASME, Boiler and Pressure Vessel Code,Section XI, Inservict Inspection, Article IWV-3400.

!. FSAA , S<c,j19q /'-II~ 2.

/i'~, S:,;.,~ N. f4 S?&::.11=1~ ~\GtfJli.&..S 'E .<.. 1 ~~fol\ C.E:t.l~TC& l..Dl..J ~~u2'E:- At.Jl::I' co~~"'"'~li.Ni- IJ.\C:a'4. P~t:'S,~£!:') A.Cllr ibTE> vtJT:)&tit. ~o""

'S.3 I 11

.!.IJ~~u~*~~t.J.

CEOG STS B 3.7-17 Rev 1, 04/07/95 Revised

03/15/99

CST 8 3.7.6 BASES LCO (continued}

is being relied upon for heat removal,f the is require~

to be OPERABLE. CoNDet.JSlt:r~ TO~ AIJ'() ~'( 'I In MODES 5 and 6, the System is not required.

ACTIONS B.l and 8.2 CDA.li::e..JS-,.C-v'OLvME-Oi!l'He SAcJtvP svm If the cannot be restored to OPERABLE status within the associ ted Com letion Time the must be placed in a c:I}

0 in w c tatu the e L 4,1 oes no apply. To achieve this must be placed in at least HOOE 3 within 6 ours, and in MOOE for heat removal, within wh t reliance on steam generator hours. The allowed Completion so (continued}

CEOG STS B 3.7-34 Rev 1, 04/07/95

Revised 03/15/99

SECTION 3.7 INSERT The CCW System consists of three pumps connected in parallel to common suction and discharge headers. The discharge header splits into two parallel heat exchangers and then combines again into a common distribution header which supplies various heat loads. A common surge tank provides the necessary net positive suction head for the CCW pumps and a surge volume for the system. A train of CCW is considered to be that equipment electrically connected to a common safety bus necessary to transfer heat acquired from the various heat loads to the Service Water System (SWS). There are two CCW trains, each associated with a Safeguards Electrical Distribution Train which are described in Specification 3.8.9, "Distribution Systems - Operating." The CCW train associated with the Left Safeguards Electrical Distribution Train consists of two CCW pumps (P-52A. P-52C), both CCW heat exchangers (E-54A, E-54B), the CCW surge tank (T-3), associated piping, valves, and controls for the equipment to perform their safety function. The CCW train associated with the Right Safeguards Electrical Distribution Train consists of one CCW pump (P-52B), both CCW heat exchangers (E-54A, E-548), the CCW surge tank (T-3), associated piping, valves, and controls for the equipment to perform their safety function. The pumps and valves are automatically started upon receipt of a safety injection actuation signal and all essential valves are aligned to their post accident positions. CCW valve repositioning also occurs following receipt a recirculating actuation signal (RAS) which aligns associated valves to provide full cooling to the component cooling water heat exchangers during the recirculation phase following a design basis Loss of Coolant Accident (LOCA) .

B 3.7-36 Revised 03/15/99

SECTION 3.7 INSERT 1 Specific signals (e.g., safety injection) are tested under Section 3.3, "Instrumentation." If the isolation valve for the noncritical service water header (CV-1359) or for the contairunent air cooler VHX-4 (CV-0869) fail to close, then both trains of SWS are considered inoperable due to the diversion of cooling water flow.

INSERT 2

..... in the "with standby power available" mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, "Electrical Power Systems."

INSERT 3 This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE.

B 3.7-44 Revised 03/15/99

UHS B 3.7.9 BASES BACKGROUND cooling tow ut a make (continued)

Additional information on the design and operation of the system alon with a list of components served can be found

@ in ~SAt!!, SE"cno1..1 '1.1 ( t.~t:. I)

APPLICABLE SAFETY ANALYSES (continued)

CEOG STS B 3.7-47 Rev 1, 04/07/95 Revised 03/15/99

INSERT The minimum water level of the UHS is based on the NPSH requirements for the SWS pumps. The NPSH calculation assumes a minimum water level of 4 feet above the bottom of the pump suction bell which corresponds to an elevation of 557.25 ft. Violation of the SWS pump submergence requirement should never become a factor unless the Lake Michigan water level falls below the top of the sluice gate opening which is at elevation 568.25 ft. Early warning of a falling intake water level is provided by the intake structure level alann.

The nominal lake level is approximately 580 ft mean sea level. The minimum water temperature of the UHS is ...

B 3.7-47 Revised 03/15/99

UHS 8 3.7.9 BASES

REFERENCES 1. FSAR, Sect1on~9~~ ( ; )

j2. /RegUJ'(torVCuidt/(.27 .j fi)

3. D8D, 1.0~

z. F'SAe, 5C::cr1c~ Jy .1 B CEOG STS B 3.7-49 Rev 1, 04/07/95 Revised 03/15/99

SECTION 3.7 INSERT 1

..... isolate the safeguards rooms by closing the inlet and exhaust plenum dampers on the initiation of a high radiation alarm from their respective airborne particulate monitor. This isolation lowers the offsite dose to well within 10 CFR 100 (Ref. 1) limits if a leak should occur. Typically, high radiation would only be expected due to excessive leakage during the recirculation phase of operation following a loss of coolant accident (LOCA).

INSERT 2

..... supply plenum damper, an exhaust plenum damper, a radiation monitor, and associated piping, valves, and duct work.

INSERT 3

..... which is addressed in LCO 3.3.10, "Engineered Safeguards Room Ventilation (ESRV)

Instrumentation" INSERT 4

..... shut, isolating the affected safeguards room(s) from the rest of the auxiliary building ventilation system lowering the leakage to the environment from the auxiliary building.

B 3.7-65 Revised 03/15/99

ECCS PREACS B 3.7.13 BASES LCO ECCS PR S is consi (continued) compon ts necessa filt tion are OP An ECCS PREACS r in is considered OPERABLE when its associated~ Ar;)1A'T"10AJ MON1~~, :/\.)sr£"v"'1E-~Ar10 ,

¥Ai..l/E->. O.IVt::. t:>AMi:>e,v:.<:. ~l:&- OP~~Le:;

a. an 1

b. f 1lter strictin9 fl and are ca?, le of perf filtration f ctions; and TeAl~S AJ?

L;::_J f2:'\. _ __

APPLICABILITY In MODES l, 2, 3, and 4, the ECCS PREAC required to be EME-f'bENCr" C.o~e OPERABLE consistent with the OPERABILITY requirements of the (ECC~

C.ooc.1JJ6 5v'STl!i'M TeAIN5~

In MODES 5 and 5, the ECCS PREACS not required to be OPERABLE, since the ECCS fs not required to be OPERABLE .

ACTIONS en The day Completi c tr1bution is ss than mpletion T1 and this systet1 for th ECCS. The 7 reasonable, ased on the l (continued)

CEOG STS B 3.7-67 Rev 1, 04/07/95 Revised

03/15/99

SECTION 3.7 INSERT 1 Condition A addresses the failure of one or both ESRV Damper train(s). Operation may continue as long as action is immediately initiated to isolate the affected engineered safeguards room. With the inlet and exhaust dampers closed or if the -inlet and outlet ventilation plenums are adequately sealed, th~ engineered safeguards room is isolated and the intended safety function is achieved, since the potential pathway for radioactivity to escape to the environment from the engineered safeguards room has been minimized.

The Completion Time for this Required Action is commensurate with the imponance of maintaining the engineered safeguards room atmosphere isolated from the outside environment when the ECCS pumps are circulating primary coolant after an accident.

B 3.7-67 Revised 03/15/99

~ =.FJt.L tia..JLu'"O:l. Pk.A

'v6+1LA+1ori CZ 'f'HclV' p ~.1.11.f:. ~ ~1. ll.

FBACS @

B 3.7,li_

B 3.7 PLAHT SYST~ .

B 3.7.~Fuel lu5~hrr cWsyste111

. ,, ~ *n (FBACS)

BASES BACKGROUND @

he FBACS cons1s of two independent, redundan trains.

ich trifn consi ts of i heiter, i prefilter o de11ister, a igh efficiency partfcul&te iir (HEPA) filter, in ictivited h&rco&l idsor r section for removal of g&se us activity principally i dines), &nd i fin. Ductwork, alves or ampers, and strumentation also fona part f the system, s well as d isters, functioning to reduce the relative Ullidity of he a1r streilll. A second bank of HEPA filters ollows the adsorber section to collect c bon fines and rovide ba up in case of failure of the &in HEPA filter ink. The downstreill HEPA filter is not credited in the nalys1s, ut serves to collect charcoa fines, and to back p the u treui HEPA filter should it velop i leik. The ystet1 i 1t1ates filtered ventilation f the fuel handling uildin following receipt of a high diation signal.

ht F CS 1s a standby syste11, part f which ~Y also be per1 Id during nor111l unit operiti ns. Upon receipt of the ctu ing signal, nor111l air disch es frOll the fuel a ing building, the fuel handl

g building is isolated, nd th1 str1111 of ventilation ai~ discharges through the sy .. filter trains. The prefi ters or demisters remove larg1 particles in the air, and &ny entrained water lits pr1s1nt, to prevent e essive loading of the HEPA r and charcoal adsorber *

(J)

CJqg ST$)

Pa.'-i s 111 tN" 83.7-71 (continued)

Rev 1, 04/07/95 Revised 03/15/99

SECTION 3.7 INSERT 1 The fuel handling area is served by two separate subsystems one being part of the original plant design, and the other being added as part of the Auxiliary Building Addition.

The original plant design consists of a supply plenum and an exhaust plenum including associated ductwork, dampers, and instrumentation. The supply plenum contains one prefilter.

two heating coils, and one supply fan. The exhaust plenum contains two filter banks (normal and emergency) configured in a parallel flow arrangement, and two independent exhaust fans which i draw air from a common duct. The "normal filter bank" contains a prefilter and a High I Efficiency Particulate Air (HEPA) filter. The "emergency filter bank" contains a prefilter, HEP A I filter, and an activated charcoal filter. I I

The Auxiliary Building Addition, which was added to serve the spaces at the north end of the I spent fuel pool, also consist of a supply plenum and exhaust plenum. The supply plenum is I configured similar to the supply plenum provided in the original plant design and includes one 1*

prefilter, two heating coils, and one supply fan. The exhaust plenum is different from the I original plant design in that it only contains one filter bank consisting of a prefilter and HEP A I filter, and two common exhaust fans. I I

During normal plant operations, the Fuel Handling Area Ventilation System supplies filtered and I heated (as needed) outside air to the fuel handling area. The exhaust fans draw air from the fuel I handling area through the normally aligned prefilters and HEPA filters and discharge it to the I unit stack by way of the main ventilation exhaust plenum. I I

During plant evolutions when the possibility for a fuel handling accident exists, the Fuel I Handling Area Ventilation System is configured such that all fans are stopped except one exhaust I fan in the original plant subsystem aligned to the "emergency filter bank." The "normal filter I bank" in the original plant design is isolated by closing its associated inlet damper. Thus, in the I event of a fuel handling accident, the fuel handling area atmosphere will be filtered for the I removal of airborne fission products prj.or to being discharged to the outside environment. I I

B 3.7-71 Revised 03/15/99

FBACS B 3.7.14 BASES (continued)

APPLICABLE The FBACS is designed to miti ate the conse uences of a SAFETY ANALYSES fuel handlin accident in w 1c a r s 1n the fuel

, asse y a e assume to be dimaged. ~ analysis of the l fuel hand ing accident is given in Re rence 3. The Design Basis Ac dent analysis of the fuel ndling accident

! assurnes hat only one train of the F CS is functional, d i to a si gle failure that disables~e other train. The i accide analysis accounts for the reduction in airborn i radio tive 111;1terial provided by e re*aining one trai of

.I this /iltration system. The amoynt of fission product avaifable for release from the f,(lel handling building is i1 detfr'mined for a fuel handling)ccident. These assu tions

an the analysis follow the gu,,tdance provided in Re latory G de 1.25 (Ref. 4).

I e FBACS satisfies Criteri 3 of the NRC Policy

LCO I.rJS£R\ 2 Q)

The FBACS is considered OPERABLE when the individual -r~~

c~onents necessary to control ex osure in the fuet.....J handling building are OPERABLE r . ~'fBACS

~is considered OPERABLE w~en oc1ate :

Cot J.;.} Fan is !!fRAB!JV ~ I~SOCf 3*

~~r\"

b. HEPA filter and charcoal adsorber are not excessively

c. .m restricting flow, and are capable of performing their filtration functions; and a&1iterJ /uctwork, ~and dampers are lE, and air circulatio~ maintained.

APPLICABILITY In MODES l, 2, 3, and 4, the FBACS is~uired to be OPERABLE to P, ovide fissjon product re val associated wi ECCS leaks e to a LOCA (refer to LC 3.7.13, *Emergenc

--~

(

(continued)

CEOG STS B 3.7-72 Rev 1, 04/07/95 Revised 03/15/99

SECTION 3.7 INSERT 1

... or fuel cask drop accident by limiting the amount of airborne radioactive material discharged to the outside atmosphere.

The results and major assumptions used in the analysis of the fuel handling accident are presented in FSAR Section 14.19. For the purpose of defining the upper limit of the radiological consequences of .a fuel handling accident, it is assumed that a fuel bundle is dropped during fuel handling activities and all the fuel rods in the equivalent of an entire assembly (216) fail. The bounding fuel handling accident is assumed to occur in containment two days after shutdown. No containment isolation is assumed to occur. As such, the released fission products escape to the environment with no credit for filtration. The results of this analysis have shown that the offsite doses resulting from this event are within the guideline of 10 CFR 100. In the event a fuel handling accident were to occur in the fuel handling area, the radioactive release would pass through the "emergency filter bank" significantly reducing the amount of radioactive material released to the environment. Thus, the consequences of a fuel handling accident in the fuel handling area are deemed acceptable with or without the "emergency filter bank" in operation since they are no more severe than the consequences of a fuel handling accident in containment .

The results and major assumptions used in the analysis of the fuel cask drop accident are presented in FSAR Section 14.11. For the purpose of defining the upper limit of the radiological consequences of a fuel cask drop accident, it is assumed that all 73 fuel assemblies in a 7 x 11 Westinghouse spent fuel pool rack with a minimum decay of 30 days are damaged and release their fuel rod gap inventories. Three fuel cask drop scenarios were analyzed to encompass all fuel cask drop events. They are:

1. A fuel cask drop onto 30 day decayed fuel with the Fuel Handling Area Ventilation System aligned for emergency filtration with a conservative amount of unfiltered leakage. All isolatable unfiltered leak path are assumed to be isolated prior to event initiation.

- 2. A fuel cask drop onto 30 day decayed fuel with the Fuel Handling Area Ventilation System aligned for emergency filtration with a conservative amount of unfiltered leakage. This scenario determined the maximum amount of non-isolatable unfiltered leakage than can exist and still meet offsite dose limits. This scenario also assumes isolation of isolable leak paths prior to event initiation.

3. A fuel cask drop onto 90 day decayed* fuel without the Fuel Handling Area Ventilation System aligned for emergency filtration. This scenario needs no assumptions as to unfiltered leakage or post-accident unfiltered leak path isolation times since all radiation is assumed to be released unfiltered from the fuel handling area.

B 3.7-72 Revised

SECTION 3.7 INSERT 1 Con't The results of the analysis show that the radiological consequences of a fuel cask drop in the spent fuel pool meet the acceptance criteria of Regulatory Guide 1.25 (Ref. 4) and NUREG-0800 Section 15.7.5 (Ref. 5) for all scenarios. In addition, the dose from all scenarios are less than 25 % of the dose guidelines in 10 CFR 100. For scenario 2, the analysis shows that a maximum of 20 % charcoal filter bypass from non-isolatable leak paths can be accommodated while still meeting 253 of the 10 CFR 100 guidelines.

Filtration of the fuel handling area atmosphere following a fuel handling accident is not necessary to maintain the offsite doses within the guidelines. of 10 CFR 100. Thus, a total system failure would not impact the margin of safety as described in the safety analysis.

However, analysis has shown that post accident filtration by the Fuel Handling Area Ventilation System provides significant reduction in offsite doses by limiting the release of airborne radioactivity. Therefore, for the fuel handling accident, the Fuel Handling Area Ventilation System satisfies Criterion 4 of 10 CFR 50.36(c)(2).

Filtration of the fuel handling area atmosphere following a fuel cask drop on irradiated fuel assemblies with < 90 days decay is required to maintain the offsite doses within the guidelines of 10 CFR 100. Therefore, for the fuel cask drop accident, the Fuel Handling Area Ventilation System satisfies Criterion 3 of 10 CFR 50.36(c)(2).

INSERT 2 The LCO for the Fuel Handling Area Ventilation System ensures filtration of the fuel handling area atmosphere is irnme~iately available in the event of a fuel handling accident, or a fuel cask drop accident. As such, the LCO requires the Fuel Handling Area Ventilation System to be OPERABLE with one fuel handling area exhaust fan, aligned to the "emergency filter bank",

in operation.

INSERT 3

... aligned to the "emergency filter bank" and in operation to ensure the air discharged to the main ventilation exhaust plenum has been filtered. Operation of only one fuel handling area exhaust fan ensures the design flow rate of the "emergency filter bank" is not exceeded.

B 3.7-72 Revised

SECTION 3.7 INSERT 4 Inclusive to the requirement to align the "emergency filter bank" is that the "normal filter bank" is isolated by its associated inlet damper to prevent the release of unfilter air.

INSERT 5 The Fuel Handling Area Ventilation System must be Operable, aligned, and in operation whenever the potential exists for an accident that results in the release of radioactive material to the fuel handling area atmosphere that could exceed previously approved offsite dose limits if released unfiltered to the outside atmosphere. As such, the Fuel Handling Area Ventilation System is required; during movement of irradiated fuel assemblies in the fuel handling building when irradiated fuel assemblies with < 30 days decay time are in the fuel handling building; during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in containment when irradiate fuel assemblies with < 30 days decay time are in the containment with the equipment hatch open, and during movement of a fuel cask in or over the spent fuel pool when irradiated fuel assemblies with < 90 days are in the fuel handling building.

The requirement for the Fuel Handling Area Ventilation System does not apply during movement of irradiated fuel assemblies or CORE ALTERATIO NS when all irradiated fuel assemblies in the fuel handling building, or all irradiated fuel assemblies in the containment with the equipment hatch open, have decayed for 30 days or greater since the dose consequences from a fuel handling accident would be of the same magnitude without the filters operating as the dose consequences would be with the filters operating and two days decay. In addition, the requirement for the Fuel Handling Area Ventilation System does not apply during fuel cask movement when all irradiated fuel assemblies in the fuel handling building have decayed 90 days or greater since the dose consequences remain less than 25 3 of the guidelines of 10 CPR 100 .

B 3.7-72 Revised 03/15/99

FBACS B 3.7.14 BASES APPLICABILITY Core Cooling Sy em (ECCS) Pump Roo Exhaust Air Cleanup (continued) System (PREACS ) for units that u this system as part their ECCS PR CS.

During move nt of irradiated f el assemblies in the fue building, t e FBACS is require to be OPERABLE to mitig e the conseq ences of a fuel ha ~ling accident.

In MODES and 6, the FBACS is not required to be OP since t ECCS is not req red to be OPERABLE.

ACTIONS A. l 1.rd A.?,

If one FBACS train i inoperable, action must be taken to restore OPERABLE st tus within 7 days. During this time period, the remain' g OPERABLE train is adequate to perf. rm the FBACS function The 7 day Completion Ti11e is l reasonable, based on the risk from an event occurring 't requiring the in erable FBACS train, and ability of remaining FBACS rain to provide the required protec I B.! and B.Z ~ ii l

In MOOE 1, 2 3, or 4, when Required Action A.I c:;nnot be completed wi hin the Completion Time, or when bo h FBACS trains are noperable, the unit must be placed n a*HOOE in which the 0 does not apply. To achieve this status, the unit must placed in MOOE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , nd in MOOE 5 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . The allowed Completion Ti are reasonab 1, based on operating experience, reach the requir unit conditions from full power co ditions in an orderl *inner and without challenging uni systems.

Wh Required Action A.I cannot be co leted within the re uired Completion Ti111e during mov nt of irradiated fuel i' the fuel building, the OPERABLE F CS train must be A

arted i111111diately or fuel movemen suspended. This action sures that the remaining train i OPERABLE, that no detected failures preven~ing sy em operation will occur, nd that any active failure will e readily detected .

CEOG STS B 3.7-73 (continued)

Rev 1, 04/07/95 Revised 03/15/99

SECTION 3.7 INSERT 1 If the Fuel Handling Area Ventilation System is not aligned to the "emergency filter bank", or one exhaust fan is not in operation, or the system is inoperable for any reason, action must be taken to place the unit in a condition in which the LCO does not apply. Therefore, activities involving the movement of irradiated fuel assemblies and CORE AL TERATIONS and movement of a fuel cask in or over the spent fuel pool must be suspended immediately to minimize the potential for a fuel handling accident.

The suspension of fuel movement and CORE AL TERATIONS, and fuel cask movement shall not preclude the completion of placing a fuel assembly, core component, or fuel cask in a safe position.

B 3.7-73 Revised 03/15/99

FBACS B 3.7.14 BASES ACTIONS If the.system is ot placed in operation, this action requires suspens on of fuel movement, which precludes a fuel handling accide t. This does not preclude the movemen of fuel to a safe position.

When ains of the FBACS are inoperable duri of irrad ted fuel assemblies in the fuel build g, action.

must be aken to place the unit in a condition n which the LCO do not apply. This LCO involves inmed1 ely suspe ing movement of irradiated fuel assem ies in the fuel uilding. This does not preclude the vement of fuel to safe position .

SURVEILLANCE REQUIREMENTS SR 3.7.14.1 Standby syste s should be checked periodically to that they fu ction properly. As the environment operating c ditions on this system are not seve , testing each train nee every month provides an adequat check on this syst . Monthly heater operation dries o any moisture ccU11Ul1ted in the charcoal from hum* ity in the lllbi*nt ir. [Systems with heaters must be erated for

~ 10 c tinuous hours with the heaters ener zed. Systems witho h*1t1rs need only be operated for 15 minutes to dellO trate the function of the syst111.] he 31 day Fre ency is based on the known reliabil y of the equipment an the two train redundanc av *

3

7

14 ClJ This SR verifies the performance of FBACS filter testing in CD a rdance with the~entilation Filter Testing Program Y . The FBACS filter tests are in accordance with the Requ atory Guide 1.52 (Ref:lAJ. The [VFTPJ 1nc1aaes testing HEPA filter perfonnance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal {gener.al use and following specific

(

(continued)

CEOG STS B 3.7-74 Rev 1, 04/07/95 Revised 03/15/99

FBACS B 3.7.14 BASES I

SURVEILLANCE SR 3 . 7 .14 ~ (continued)

REQUIREMENTS oper*ations). Specific test frequencies and additional information are discussed in detail in the,,X'VFTP~

~

~ -

2.

SR 3.7.14.~

the FBACS filter bypass amper is necessary to ensure hat the system functions roperly. The OPERABILITY of th FBACS filter bypass damp is verified if it can be clos

The 18 month Frequenc is consistent with that spe ified in Reference 6.

(continued)

CEOG STS B 3.7-75 Rev 1, 04/07/95 Revised 03/15/99

SECTION 3.7 INSERT 1 Fuel Handling Area Ventilation System has not degraded and is operating as assumed in the /

safety analysis. The flow rate /

l I

I INSERT 2 I I

When aligned to the "emergency filter bank", the Fuel Handling Area Ventilation System is I designed to reduce the amount of unfiltered leakage from the fuel handling building which, in I the event of a fuel handling accident, lowers the dose at the site boundary to well within the /

guidelines of 10 CFR 100. I I

I I

INSERT 3 I

. . . . .lower the dose to these levels . . . . . I I

1 .

B 3.7-75 Revised 03/15/QQ

FBACS B 3.7.14 BASES (continued)

REFERENCES 2.

3.

4.

Regulatory Guide 1.52., ~1~n, 1Cs~, Q.nd llb.mCl\an<..<.. ~.U10.

5 (§. NUREG-0800, Sectio11 F-sf, July l~l.J f.r fls.t ~ ~Mu.rJ-. $-1~*

liA~ ~hclt. l!. lt.c.. ... "'

S'1'S-ltnl A.r ,-, -f1ll..\-~ o.rJ

~1£:1,51 SP,t.H f'Vrl i'.3411: DM ~dvJ:/,

l'l03rlf,w Un 11s o ~ l*~-U.>>1 .

C4o I~ NUC..(CQt ~w,r Plc.1rf5., .

CEOG STS B 3.7-76 Rev 1, 04/07/95 Revised 03/15/99

Spent Fuel Assembly Stora9.1.

B 3.7.Xi)

B *3. 7 PLAHT SYSTEMS

(CA f\ 1~~ rj.).-~ffA::ii9~~~r;r-::

<1NSE:RT')

APPLICABLE The spent fuel storage facility is designed for SAFETY ANALYSES noncriticality by use of adequate spacing, and *flux trap*

construction whereby the fuel assemblies are inserted into neutron absorbing stainless steel cans.

(continued)

CEOG STS B 3.7-89 Rev l, 04/07/95

- .\.

Revised 03/15/99

BASES (cont;nued)

APPlICABILITY ACTIONS indicating th1t rnov ng rra a ue ass e n LCO 3.0.3 would not specify any actton. If movtng irradtatld fuel asslllbl1es while in MOOE 1, Z, 3, or 4, the fuel mov ...nt ts independent of reactor operatton.

Therefor*, tn 1tther case, ;nabtltty to 110ve fuel asslllblt1s ts not suffict1nt reason to require a reactor shutdcnnt.

SURVE I LLAHCE REQUIR£MEHTS G) rJ) i'"Al.I&

REFERENCES None.

CEOG STS B 3.7-90 Rev 1, 04/07/95 Revised 03/15/99

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.7.14, FUEL BUILDING AIR CLEANUP SYSTEM (FBACS)

Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARK.UPS. " The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.

1. The brackets have been removed and the proper plant specific information or value has been provided.

2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.

3. The requirement/statement has been deleted since it is not applicable to this facility.

The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5. This change reflects the current licensing basis/technical specification. The design of the Fuel Handling Area Ventilation System is such that there is only one "train" not two as NUREG-1432 describes. The cleanup portion of the system contains two filter banks (normal and emergency) configured in a parallel flow arrangement, and two independent exhaust fans drawing air from a common duct. The "normal filter bank" contains a prefilter and a High Efficiency Particulate Air (HEPA) filter. The "emergency filter bank" contains a prefilter, HEPA filter, and an activated charcoal filter. As such, all statements concerning two trains, the reference to the second train, or single failure proof are deleted.

Palisades Nuclear Plant Page 1of4 03/15/99

AITACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.7.14, FUEL BUILDING AIR CLEANUP SYSTEM (FBACS)

Change Discussion

6. The Applicability of ISTS 3. 7 .14 has been revised to match the plant conditions when the potential for a fuel handling accident exist. This includes Core Alterations, or movement of any fuel assembly in the containment, when irradiated fuel assemblies with < 30 days decay time are in the containment and the equipment hatch is opened.

With the equipment hatch opened, the containment atmosphere is in direct contact with the fuel handling building atmosphere. In the event of a fuel handling accident in containment, the Fuel Handling Area Ventilation system is capable of filtering the airborne radioactive material in the containment atmosphere prior to being released to the outside atmosphere. In addition, the Applicability also includes movement of a fuel cask in or over the spent fuel pool. The fuel cask drop accident is presented in FSAR Section 14.11 and forms the bounding heavy load accident involving damage to stored irradiated fuel in the spent fuel pool. Conforming changes have been made to the Bases.

7. The Actions of ISTS 3. 7 .14 have been revised to address the most probable causes for failure to meet the requirements of proposed LCO 3.7.12. As such, the Actions address the conditions when the Fuel Handling Area Ventilation system is inoperable, not properly aligned, or not in operation. Since the Fuel Handling Area Ventilation system consists of a single train aligned in its accident mitigation mode, the only appropriate Required Actions upon failure to meet the LCO is to immediately suspend fuel handling, CORE ALTERATIONS, and fuel cask movement activities.

Conforming changes have been made to the Bases.

8. ISTS SR 3.7.14.1 requires each FBAC train be operated for ~10 hours (for plants with heaters), or ~ 15 minutes (for plants without heaters) every 31 days. The intent of this SR is to ensure the standby FBAC system functions properly. For plants that rely on automatic actuation signals, or whose Applicability includes Modes 1, 2, 3, and 4, performance of this SR fulfill the intended function. However, for the Palisades plant, the Fuel Handling Area Ventilation system is required to be in operation whenever the plant is in the condition specified in the Applicability. Since SRs are only required to be met during the condition specified in the Applicability, performance of a system functional test would be redundant to the requirements of the LCO. Therefore, specifying this SR in the ITS is not necessary.

Palisades Nuclear Plant Page 2 of 4 03/15/99 I

I I

J

Chanu SPECIFICATION 3.7.14, FUEL BUILDING AIR CLEANUP SYSTEM (FBACS)

Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS

9. This change reflects current plant design. The Fuel Handling Area Ventilation system does not include an automatic actuation feature. The system is manually configured in the emergency filtration mode prior to entering the conditions specified in the Applicability. As such, the tests required by ISTS 3.7.14.3 and ISTS 3.7.14.5 are not applicable and have not been included in the ITS.

10. ISTS SR 3.7.14.5 has been changed to reflect current analysis assumptions and methods of performance that prove the Fuel Handling Area Ventilation system is operating as required. The analysis that credits the Fuel Handling Area Ventilation System assumes a flow rate of 7300 cfm +I- 20 %. No specific assumptions are made to the internal pressure of the fuel handling building relative to atmospheric pressure. Performance of this test on a Staggered Test Basis is not applicable since the Fuel Handling Area Ventilation system consists of a single train.

11. The Bases Background section for ISTS 3. 7 .14 has been revised to reflect the design of the Palisade's Fuel Handling Area Ventilation system. The level of detail provided in the revised Bases is comparable to the level of detail provided in the ISTS.

12. The Bases Applicable Safety Analyses section for ISTS 3. 7 .14 has revised to reflect specific plant analyses. The fuel handling accident in the fuel handling area is bounded by the fuel handling accident in containment which assumes all fission products release to the containment atmosphere are released to the outside environment .with no credit for filtration. The NRC has previously concluded that the consequences of a fuel handling accident in the spent fuel area are acceptable with or without the charcoal filters operating. As part of Amendment 81 to the Palisade's Technical Specifications the NRC stated "the dose with the filter system operating was calculated to be 9 rem to the thyroid. If the filtration system was not operating, the dose would have been 91 rem which is still appropriately within the guidelines of 10 CFR 100 (i.e.,

< 100 rem thyroid)." Since operation of the Fuel Handling Area Ventilation System is not part of a primary success path that functions to mitigate a design basis accident, but instead, has been shown to be significant to public health and safety, the criterion satisfied in 10 CFR 50.36 for the fuel handling accident has been stated as Criterion 4 .

Palisades Nuclear Plant Page 3 of 4 03/15/99

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.7.14, FUEL BUILDING AIR CLEANUP SYSTEM (FBACS)

Discussion 12 (continued)

The fuel cask drop accide_nt forms the basis for a heavy load drop in the spent fuel pool that results in damage to stored irradiated fuel assemblies. This analysis was performed to support storage of spent fuel assemblies in the Independent Spent Fuel Storage Installation and is discussed in FSAR Section 14.11. The analysis shows acceptable radiological consequences when crediting filtration by the Fuel Handling Area Ventilation System for certain scenarios. Since operation o_f the Fuel Handling Area Ventilation System is part of a primary success path that functio~ to mitigate a design basis accident, the criterion satisfied in 10 CFR 50.36 for the cask drop accident has been stated as Criterion 3.

Palisades Nuclear Plant Page 4 of 4 03/15/99

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.7.18, SPENT FUEL ASSE:MBLY STORAGE Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.

1. The brackets have been removed and the proper plant specific information or value has been provided.

2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.

3. The requirement/statement has been deleted since it is not applicable to this facility.

The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5. This change reflects the current licensing basis/technical specification.

6. The storage of failed fuel is accomplished by the use of canisters that fit in the same storage racks as the fuel assemblies themselves. Therefore, the storage pool does not have any specifically designed rack(s) for failed fuel. The reference to a specific number of storage locations for failed fuel is deleted .

Palisades Nuclear Plant Page 1of2 03/15/99

A'IT ACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.7.18, SPENT FUEL ASSEl\.fBLY STORAGE Chanw: Discussion

7. ISTS 3. 7 .18 applies to plants which restrict the storage of fuel assemblies in high density storage locations based on meeting an acceptable combination of initial enrichment and discharge bumup. For fuel assemblies which do not meet the initial enrichment and discharge bumup requirements, the assemblies may be stored in compliance with other NRC approved methods or configurations as stipulated in ISTS 4.3. l. l. ISTS SR 3. 7 .18.1 requires an administrative verification of the initial enrichment and discharge bumup of a fuel assembly prior to storing any assembly in a Region 2 location. For the Palisades Plant, storage of fuel assemblies in high density racks (Region m is only permitted for fuel assemblies which meet the initial enrichment and discharge bumup requirements. Alternate storage methods or configuratiop.s (e.g.,

checkerboading) in Region II has not been approved by the NRC. Therefore, reference to storage of fuel assemblies in accordance with Specification 4. 3 .1.1 in the LCO, SR, and SR Bases has been deleted. Assurance that fuel assembly enrichments do not exceed the limits of Region I locations (ITS 4. 3 .1.1) is controlled administratively in the design of new cores and the procurement of new fuel.

Palisades Nuclear Plant Page 2of2 03/15/99

ML18066A443

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