Text

Proposed Tech Specs Pages Re 980126 Submittal & Replacement Pages for Sections 3.3,3.4 & 3.9
	ML18066A501
Person / Time
Site:	Palisades
Issue date:	06/11/1999
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
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ENCLOSURE 2 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 NRC COMMENTS CONCERNING ITS SECTION 3.3 JANUARY 26, 1998 ITS SECTION 3.3 MARKED TO SHOW CHANGES

~

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

9906 100110 qqoo 11 PDR ADOCK 05000255 p PDR

- ----,1

Definitions 1.1

" 1.1 Definitions CHANNEL CALIBRATION all devices in the channel required for chonnel OPERABILITY ond * .

(

Whenever a RTO or thermocouple Calibration of instrument channels with Resistance sensing element is replaced, the Temperature Detector (RTD) or thermocouple sensors 'DEF' next required CHANNEL CALIBRATION may consist of an inplace qualitative assessment RA!

sholl include on inplace cross of sensor behavior and normal calibration of the calibrotion thot compares the other remaining adjustable devices in the channel.

sensing elements with the recently

. ins tolled sensing element.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative as~essment, by observation, of channel behavior during operation. This determination shall include, where.possible, comparison of the channel indication and status to other indications or.

status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a.

"------b. Digital channels-the use of diagnostic 'DEF' of oll devices in the chonnel programs to test di g;ta l hardware and the RAI required for channel OPERABILITY. injection of simulated process data into the channel to verif OPERABILITY, lu ng la ip Palisades Nuclear Plant 1.1-2 Amendment No. 01/20/98

RPS Instrumentation

3. 3. 1

3.3 3.3.1 INSTRUMENTATION Reactor Protective System (RPS) Instrumentation Cl-\i\.t.l.\.le.LS, A.1-lD ASSOC..lA"Tet:> 1..~Ro Pow~ Mooe- cPM LCO 3.3.1 Four RPS trip units,~ associated instrument~ Bypass removal channels for each Function in Table 3.3.1-1 shall be OPERABLE.

APPLICAB IL ITV:

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME

A. --------NOTE---------- A.1 Not applicable to High.

Startup Rate,~ Loss of Load Functions.

~

Place affected trip unit in trip.

~ o~ C.1-'N\ SVPA~ R&MO\/A.L) 7 days One or more Functions '!.

with one RPS trip unit or associated instrument channel inoperable .

Palisades Nuclear Plant 3.3.1-1 Amendment No. 01/20/98

-- -- - -~on:-- - - - --

t-!OT ~?'?L\C.A6LE- -ro ~'""µ,. RPS Instrumentation

ACTIONS CONDITION STA.RTuP RA.IE:- oR, Los~

Ol=o LOC...D HJ~'-llO~S.

REQUIRED ACTION

3. 3 .1 COMPLETION TIME i<AI.

3.'3.t - I One or more Functions 3.~.1-lo with two RPS trip units or associated instrument channels Restore one trip unit 7 days inoperable. and associated instrument channel to OPERABLE status.

Two power range Restrict THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months channels inoperable. POWER to ~ 70% RTP .

D.

c+'~ME.D\~TC1.."I' )

IY~utl 0.2 One High Startup Rat~ Restore trip unit and. Prior to ~

Trip unit or associated instrument entering m~

associated instrument channel to OPERABLE from MODE~ g~

channel inoperable. status. "5.l.1-1 One Loss of Load trip Restore trip unit and unit or associated associated instrument instrument channel channel to OPERABLE

inoperable.

Palisades Nuclear Plant status.

3.3.1-2 Amendment No. 01/20/98

RPS Instrumentation

ACTIONS CONDIT ION REQUIRED ACT ION

3. 3 .1 COMP LET ION TIME

RAI.

3,3.1-1

~

Required Action and Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

£.s,~ ED OR Verif no more than 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months one control rod is Control room ambient capable of being air temperature withdrawn.

> 90°F.

OR Verify PCS boron 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months concentration is at REFUELING BORON CONCENTRATION .

Palisades Nuclear Plant 3.3.1-3 Amendment No. 01/20/98

RPS Instrumentation

3. 3 .1 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1-1 to detennine which SR shall be perfonned for each Function.

SURVEILLANCE FPC:QUENCY SR 3.3.1.l Perfonn a CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.2 Verify control room temperature is ~ 90°F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.3 -------------------NOTE----------------~---

Not required to be perfonned until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is ~ 15% RTP.

"E.~C.O 'Ze:.

Perfonn calibration (hea balance only) and adjust the power range and ~T power channels to agree with calorimetric calculation if the absolute difference is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months I

~

~A:t.

3."3.\-4 SR 3.3.1.4 31 days SR 3.3.1.5 92 days SR 3.3.1.6 range excore channels 92 days Palisades Nuclear Plant 3.3.1-4 Amendment No. 01/20/98

RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.7 Perfonn a CHANNEL FUNCTIONAL TEST of High Once within Startup Rate and Loss of Load Functions. 7 days prior to each reactor startup SR 3.3.1.8 18 months

*-*** ~oTE:..-**--*-** - - -**

~£:.\)T'Ro...i 'OETE"-'Tl:lli:S ARe. '6:.~LL)'t>E.1:> 'F~Ot--\

T~ C~AUt-.le:t. CAL1i3e'Al\o~ .

Palisades Nuclear Plant 3.3.1-5 Amendment No. 01/20/98

RPS Instrumentation

3. 3.:

Table 3.3.1-1 (page 1 of 2)

Reactor Protective System Instrumentation A'PPL \C..A.3!£- SURVEILLANCE FUHC TION ~ /'.Q~:::.:~ REQUIREMENTS ALLOWABLE VALUE

1. Variable High Power Trip SR 3. 3 .1.1 ~ 15% RTP above current SR 3.3.1.2 THERMAL POWER with a SR 3.3.1.3 minimum of > 30% RTP SR 3.3.1.4 and a maximum of s SR 3.3.1.5 106.5% RTP SR 3. 3 .1. 6 SR 3.3.1.8

2. High Startup Rate~-;:.-=-/*;,. SR 3.3.1.1 NA SR 3.3.1.7 SR 3.3.1.8 R;a..J:

.,.*:;_; fci..) r-:i:_; ~.3.\-1

3. Low Primary Coolant Syste11 Flow~ 1 :"2.,_, 1 ~- , S SR 3.3.1.1 95%

....ee 1r.;

SR 3.3.1.5 SR 3.3.1.8

4. Low Stea* Generator A Level ~:? SR 3. 3 .1.1 < 25.9% *narrow r'ange SR 3. 3 .1. 5 SR 3.3.1.8

. .., '2(~ . .(<;1.'; .,,.,~'

5. Low Stea. Generator B Level 7~.!? **'-i-=- 1 '-'. , '.J ** ' SR 3. 3 .1.1 25.9% narrow range SR 3. 3 .1. 5 SR 3.3.1.8 SR 3. 3 .1.1 ~ 500 psi a SR 3.3.1.5 SR 3. 3 .1. 8 7.

Mi Low Stea. Generator B Pressurellil:,

I

'1 '2, '5 (a_\ *-J 1

~; ~ ~o_)

, SR 3. 3 .1.1 < 50Q psi a I

. T~\p'.C: SR 3. 3 .1. 5 SR 3. 3 .1. 8

8. High Pressurizer Pressure-:-~"? SR 3.3.1.1 s 2255 psia SR 3.3.1.5 SR 3. 3 .1. 8 Bypass shall

Palisades Nuclear Plant 3.3.1-6 Amendment No. 01/20/98

l RPS Instrumentation

*Table 3.3.1-1 (page 2 of 2)

Reactor Protective System Instrumentation 3.3.1 f..7.--'1_* CA eu..:- SURVEILLANCE FUNCTION 't\~t;E-'C., REQUIREMENTS ALLOWABLE VALUE

9. Ther111al Margin/Low Pressureli!I z.~ :i*: '-1,.:i.'; -S ::;.. :. SR 3.3.l.l Table 3.3.1-2

-~ ~ ~:*. SR 3.3.l.2 SR 3.3.l.3 SR 3. 3. l. 4 SR 3.3.l.5 R.b.-1:

SR 3. 3. l. 6

'.3.1-1*

SR 3.3.l.8

10. Loss of Loadra'! Te* :7 SR 3.3.l.7 NA SR 3.3.l.8 SR 3.3.l.5 ~ 3.70 psig SR 3.3.l.8 Bypass shall E:D

\.-------*-*-------------------*-----.

(o.) ( A"DD FoO\~o\E:.. (~) FeoM F'g>Evious ?A~>

'n THERMAL POWER i~l7% RTP. le)l(a~

eAI 3,3'.1- I

~.~.1-lP

Palisades Nuclear Plant 3.3.1-7 Amendment No. 01/20/98

RPS Instrumentation

3. 3 .1 Table 3.3.1-2 (page 1 of 1)

Thennal Margin/Low Pressure Function Allowable Value The Allowable Value for the Thennal Margin/Low Pressure Trip, P,., 0 , is the higher of two values, P0 .,

and P,.,, both in psia:

P = 1750 0

P,., = 2012(QA) (QR 1) + 17 .O(T 1,) - 9493 Where:

QA = - 0.720(ASI) + 1.028;* when - 0.628 ~ ASI < - 0.100 QA= - 0.333(ASI) + 1.067; when - 0.100 ~ ASI < + 0.200 QA = + 0.375(ASI) + 0.925; when + 0.200 ~ ASI s + 0.565 ASI = Measured ASI when Q ~ 0.0625 ASI = a.a when Q < 0.0625 QR 1 = 0.412(Q) + 0.588; when Q ~ 1.0 QR1 = Q; when Q > 1.0

Q = THERMAL POWER/RATED THERMAL POWER T,, = Maximum primary coolant inlet temperature, in °F AS[, T1,, and Q are the existing values as measured by the associated instrument channel.

Palisades Nuclear Plant 3.3.1-8 Amendment No. 01/20/98

RPS Logic and Trip Initiation 3.3.2 3.3 INSTRUMENTATION 3.3.2 Reactor Protective System (RPS) Logic and Trip initiation LCO 3.3.2 Six channels of RPS Matrix Logic, four channels of RPS Trip Initiation Logic, and two channels of RPS Manual Trip shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, ('FULL:-LENC;,~ Et/

MODES 3, 4, and 5, with more than oneJZontrol rod capable of being withdrawn and Primary Coolant System (PCS) boron concentration less than REFUELING BORON CONCENTRATION.

ACTIONS CONDITION REQUIRED. ACTION COMPLETION TIME

A. One Matrix Logic channel inoperable.

B. One channel of Trip A.1 B~ 1 Restore channel to OPERABLE status.

De-energize the 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 1 hour Initiation Logic affected clutch power inoperable. supplies.

c. One channel of Manual C.l Restore channel to Prior to Trip inoperable. OPERABLE status. entering MODE 2 from MODE 3 Palisades Nuclear Plant 3.3.2-1 Amendment No. 01/20/98

RPS Logic and Trip Initiation 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Two channels of Trip 0.1 De-energize the Immediately Initiation Logic affected clutch power affecting the same supplies.

trip leg inoperable.

E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met. AND Cl=ULL:" Lt.lbn1)

OR E. 2.1 (_ve_ri fy no more than 6 hours one control rod is One or more Functions capable of being with two or more withdrawn.

Manual Trip, Matrix Logic or Trip OR Initiation Logic

~

channels inoperable E.2.2 Verify PCS boron 6 hours for reasons other than concentration is at Condition D. REFUELING BORON CONCENTRATION.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1 Perfonn a CHANNEL FUNCTIONAL TEST on each 92 days RPS Matrix Logic channel and each RPS Trip Initiation Logic channel.

SR 3.3.2.2 Perfonn a CHANNEL FUNCTIONAL TEST on each Once within RPS Manual Trip channel. 7 days prior to each reactor startup

Palisades Nuclear Plant 3.3.2-2 Amendment No. 01/20/98

ESF Instrumentation 3.3.3 3.3 INS TRUMENTA TI ON 3.3.3 Engineered Safety Features (ESF) Instrumentation LCO 3.3.3 Four ESF bistables and associated instrument channels for each Function in Table 3.3.3-1 shall be OPERABLE.

APPLICABILITY: As specified in Table 3.3.3-1.

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME

A. --------NOTE----------

,~~~RA~

One or more Functions A.1 Place affected bistable in trip.

7 days with one ESF bistable or associated instrument channel inoperable.

B. -------------NOTE-~----------

LCO 3.0.4 is not applicable.

B.1 Place one bistable in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months trip.

One or more Functions with two ESF bistables AND A.Nt:> ASSOCtA.'Te-D

\ "1'£>T R.U N'.f:-"'11 CMA.tJNE or associated instrument channels B.2 Restore one bistable 7 days inoperable . 'to OPERABL~ status.

Palisades Nuclear Plant 3.3.3-1 Amendment No. 01/20/98

ESF Instrumentation 3.3.3 ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME

c. C.1 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months e c anne ino erable. j:?AL.

~\S\A.6\-'E:' AND o~ ASSOC:. I A.Tel:> 3.3.3-l I f..!C::,TIC!U~E-~I C.2 Restore channel to 7 days OPERABLE status.

BISTABLE: AJJD AS':l'.XIAT'e'P llJSTRIJ AJI D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time not met for AND

@'nction@,

. 2, -;,

\ifll or D.2 Be in MODE 4. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months

E. Required Action and associated Completion Time not met for Functions~or~

E.1 AND E.2 Be in MODE 3.

Be in MODE 5.

6 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.3-1 to determine which SR shall be performed for each Function.

SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform a CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.3.2 Perform a CHANNEL FUNCTIONAL TEST. 92 days

SR 3.3.3.3 Perform a CHANNEL CALIBRATION. 18 months Amendment No. 01/20/98 Palisades Nuclear Plant 3.3.3-2

ESF Instrumentation 3.3.3 FUNCTION Table 3.3.3-1 (page 1 of 1)

Engineered Safety Features Instrumentation 1

(APPl..lCA'Sl.E-) SURVEILLANCE MOtlES RfQU IREMENTS AlLOWAILE VALUE

l. Se f1ty Inj 1ct ion Signal (SIS)

a. Pressurizer Low Pressure 1,2,3 SR 3.J.3.1 ' 1593 psh SR 3.3.3.2 SR 3.3.3.3 Cont1i1111ent High Pressure~(CHP)

~ a. Cont1i1111ent High Pressure - Left Train 1,2,3,4 SR 3.3.3.2 ' 3 *7 psi g and SR 3.3.3.3 4~sig

b. Cont1in11ent High Pressure - Right Train 1,2,3,4 SR 3.3.3.2 l 3.7 psig ind SR J.3.3.3 s 4~sig

~

Conta11111ent High Radiation Signal (CHR)

a. Cont1i1111ent High Radiation 1,2,3,4 SR 3.3.3.l s'20 R/hour SR 3.3.3.2 SR 3.3.3.3 RAJ:

3.3.~-I St11* 61ner1tor Low Pressure Signal (SGLP)

e. St11* 6tntrator A Low Pr1ssur1 SR 3.3.J.l l 500 psh SR 3.3.J.2 SR J.3.J.3

b. Ste .. 6tnerator B Low Pressure SR 3.3.3.l 2 500 .psi1 SR 3.3.J.2 SR J.J.J.J Recirculation Actuation Signal (RAS)

1. SIRWT Low Level 1,2,3 SR 3.3.J.3 2 21 inches and

~ 27 inches 1bov1 tlnk bottom Auxilhry F11dw1ter Actuation Signal (AFAS)

a. St11* 61n1r1tor A. Low Level 1,2,J SR 3.3.3.l 2 25.9% narrow SR J.3.3.2 range SR 3.3.3.3

b. Ste.. 61n1r1tor B Low L1v1l 1,2,3 SR 3.3.3.1 l 25.9% nerrow SR 3.3.3.2 range SR 3.3.3.3 (a) Not required to be OPERABLE when all Main Stea* Isolation Valves (MSIVs) are.closed and de1ctiv1t1d, ind 111 Mein F11dw1t1r Regulating Valves (MFRVs) and MFRV bypass valves ire either closed and deactiv1t1d, or isolated by closed .. nu1l valves.

Palisades Nuclear Plant 3.3.3-3 Amendment No. -01/20/98

7. Automatic Bypass Removals a.. Pressurizer Low* Pressure Bypass 1.2 .. 3 SR 3.3.3.3 > 1700. psi a b.. Steam Generator A tow SR 3.3.3.3 > 565 psi a Press.ure Bypass SR 3.3.3.3 >565 psia c.. Steam: Generator B Low Pressure Bypass

ESF Logic and Manual Initiation

3.3.4 3.3 INSTRUMENTATION 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation LCO 3.3.4 Two ESF Manual Initiation and two ESF Actuation Logic channels and associated bypass removal channels shall be OPERABLE for each ESF Function specified in Table 3.3.4-1.

APPLICABILITY: According to Table 3.3.4-1.

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACT ION COMPLETION TIME

A. One or more Functions with one Manual Initiation. Bypass Removal, or Actuation Logic channel inoperable.

A. l Restore channel to OPERABLE status.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months B. One or more Functions B.1 Be in MODE 3. 6 hours with two Manual Initiation. Bypass AND Removal, or Actuation Logic channels inoperable for Fu~ions 1, ~

& B.2 Be in MODE 4. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months or

2 3

\

"i OR ~

Required Action and ~.3.

associated Completion Time of Condition A I no~et for ~nctions 1, ~or 2 3 Palisades Nuclear Plant 3.3.4-1 Amendment No. 01/20/98

ESF Logic and Initiation Manua~

3.3.4 ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME C. One or more Functions C.l Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with two Manual Initiation,~1~-4-~-&'4~-I AND IRlluoov#. or Actuation Logic channels C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months inoperable for

'Functions ~or I OR Required Action and associated Completion Time of Condition A not met for Functions

~.or~.

Palisades Nuclear Plant 3.3.4-2 Amendment No. 01/20/98

ESF Logic and Manual In1t1at1on

3.3.4 SURVEILLANCE FREQUENCY

( EM..H SIS A.C:.TU~T10U Cl4tl..t-l~l:::L.)

SR 3.3.4.1 Perfonn functional test ofJnonnal and 92 days (s-r~t>BV)-igfr'@nf.l power functionslusift§ te§'tl l_rc_t_.

SR 3.3.4.2

SR 3.3.4.3 Perfonn a CHANNEL FUNCTIONAL TEST.

92 days 18 months

Palisades Nuclear Plant 3.3.4-3 Amendment No. 01/20/98

ESF Logic and Manual Initiation 3.3.4

rable 3.3.4-1 (page 1 af 1)

Engineered Safety Features Actuation Logic and Manual APPLICABLE FUNCTION MODES

1. Safety Injection Signal (SIS) (a..)

1,2,3 b.

~ Containment High Pressure Signal (CHPl(eJ i

p;. 7fuatii<og~ 1,2,3,4 !

I (CHR) i i

\

Pressurizer Low Pressure 11ay be 11anually bypassed when pressurizer pressure is ~ 1700 psia. The bypass shall shall be automatically

S~LP /;£.TVA.'T\Of..l SIS Ji..C..TU"T\OtJ Palisades Nuclear Plant

'B'-1 3.3.4-4 Amendment No. 01/20/98

5EC.TION

.I:~Se\2.l" 2

'3.3 (d) Mot ~!~~-to be OPERABLE when all Main Steam Isolation l/a.lves (MSIVs} are MFflV'ldlfl::v:*1Yes NftHl'::*-*lQ$:~:

c:losd?l.M :Ckllcthated. and 111 l Hain feedwater Regulating Ya.1 ves (MFRVs.} and either closed and deactivated. or isolated by closed

~

1N$Ei?r AS F'oo\UOTE. (d~ \b r~ ~l.E:- '31"3 I y - l

DG - UV Start

3.3.5 3.3 INSTRUMENTATION 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)

LCD 3.3.5 ~anne~f Loss of Voltage Function a~~~~anne~f Degraded Voltage Function auto-initiation instrumentation DG sha 11 be OPERABLE. AfJD ASSOC.IA.TED LD6tc.. C~ANl-.lCl.S Fo'e c.Ai.C.t-+

APPLICABILITY: When associated DG is required to be OPERABLE.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Enter applicable Immediately with one channel per Conditions and DG inoperable. Required Actions for the associated DG

made inoperable by DG - UV Start instrumentation.

PEKIZ>eN\ ~ C'-\~~EL F'utJc..T1o~AL TEST I OtJ EAC.'"' Db-U\} SrAl?:r LDC.:OIC.. CHAIJNEL.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 18 months b.

Time delay:r~ 8.15 seconds at 1400 V;t.

an (~ 5,y5 sec.Ot.Jt)S ~t>Jt:> )

Palisades Nuclear Plant 3.3.5-1 Amendment No. 01/20/98 .

/

, /

/

Refuel. CHR Inyrumentati o

,/ g

,,/' /' / 3 .6 3.3.6 LCO INSTR

{3.6

/

Re/ Z -

TION Con inment

. /,

/

Hi0o~{~adiatioalfcHR)

InsLtio

- Palisades Nuclear Plant 3.3.6-1 Amendment No. 01/20/98

INSERT #3 LCO 3. 3. 6 ?~. I o;:: '2.

3.3 INSTRUMENTATION 3.3.6 Refueling Containment High Radiation (CHR) Instrumentation LCO 3.3.6 Two Refueling CHR Automatic Actuation Function channels and two CHR Manual Actuation Function channels shall be OPERABLE.

APPLICABIL1TY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME A. One or more Functions A.1 Place the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months with one channel c:hannel in trip.

inoperable.

OR

.A.2.1 Suspend CORE 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ALTERATIONS.

AND A.2.2 Suspend movement of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months irradiated fuel asemblies within containment.

B. One or more Functions 8.1 Suspend CORE Immediately with two channels ALTERATIONS

inoperable.

AND 8.2 Suspend movement of Immediately irradiated fuel assemblies within

containment .

.l_>JSE.i?\ ~3 'L(.0 7 '"2 1 *

.)* J *ID ~

r-:j, 2 O'F" I')

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.6.1 Perfonn a CHANNEL CHECK of each refueling . 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months CHR monitor channel.

SR 3.3.6.2 Perfonn a CHANNEL FUNCTIONAL TEST of each 31 days refueling CHR monitor channel.

~ (~1'-'rl.)

SR 3.3.6.~Perfonn a CHANNEL CALIBRATION of refueling 18 months CHR monitor channel.

I 3.3.6.~ Perfonn I SR a CHANNEL FUNCTIONAL TEST of each 18 months CHR Manual Initiation channel .

PAM Instrumentation

3.3.7 3.3 INSTRUMENTATION 3.3.7 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.7 The PAM instrumentation for each function in Table 3.3.7-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

NOTES------------------------------~-----

1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACT ION COMPLETION TIME

A. One or more Functions with one required channel inoperable.

A.1 Restore required channel to OPERABLE status.

30 days B. Required Action and B.1 Initiate action in Immediately associated Completion .accordance with Time of Condition A Specification 5.6.6.

not met.

c. ---------NOTE--------- C.1 Restore one channel 7 days Not applicable to to OPERABLE status.

hydrogen monitor channels.

One or more Functions with two required channels inoperable.

Palisades Nuclear Plant 3.3.7-1 Amendment No. 01/20/98

PAM Instrumentation 3.3.7 ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME D. Two hydrogen monitor D.l Restore one hydrogen 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months channels inoperable. monitor channel to OPERABLE status.

E. Required Action and E.l Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.7-1 for the or D not met. channel.

F. As required by F.

1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action E.l and referenced in AND Table 3.3.7-1.

F.2 Be in MODE 4.

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months G. As required by G.1 Initiate action in Immediately Required Action E.l accordance with and referenced in Specification 5.6.6.

Table 3.3.7-1.

Palisades Nuclear Plant 3.3.7-2 Amendment No. 01/20/98

PAM Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS.

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

These SRs apply to each PAM instrumentation Function in Table 3.3.7-1.

SURVEILLANCE FREQUENCY SR 3.3.7.1 31 days E1?

SR 3.3.7.2 Perfonn CHANNEL CALIBRATION. 18 months

~~~-~oT~~~~~~~

~e:uTeo~ DE.rec..1oec., A-eE E)£c.Lut>e:t>

t=eoM ~e. C~#\~tJE:L CAL\~Ai1otJ .

Palisades Nuclear Plant 3.3.7-3 Amendment No. 01/20/98

PAM Instrumentation 3.3.7

Table 3.3.7-1 (page 1 of 1)

Post Accident Monitoring Instrumentation CONDITIONS REFERENCED FR~

FWICTIOM REQUIRED CHAMMELS REQUIRED ACTION E.1

1. Primary Coolant System Hot Leg T~rature 2 (wide range)

2. Primary Coolant System Cold Leg T~rature 2 (wide range)

3. Wide Range Neutron Flux 2

4. Contairrnent Floor Water Level (wide range) 2

5. Subcooled Margin Monitor 2

6. Pressurizer Level (wide range) 2

7. Contairrnent Hydrogen Monitors 2

8. Condensate Storage Tank Level 2

9. Primary Coolant System Pressure (wide range) 2 F

10.

11.

12.

13.

Contairrnent Pressure (wide range)

Steam Generator A water Level (wide range)

Steam Generator B water Level (wide range)

Steam Generator A Pressure 2

2 2

2 F

14. Steam Generator B Pressure 2

15. Contairrnent Isolation Valve Position

16. Core Exit T~rature - Quadrant 4 F

17. Core Exit T~rature - Quadrant 2 4 F

18. Core Exit T~rature - Quadrant 3 4

19. Core Exit T~rature - Quadrant 4 4 F

20. Reactor Vessel Water Level 2 G

21. Contairment Area Radiation (high range) 2 G (a) Mot required for isolation valves whose associated penetration is isolated by at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured *

Palisades Nuclear Plant 3.3.7-4 Amendment No. 01/20/98

3.3 INSTRUMENT AT ION Alternate Shutdown System

3.3.8 APPLICABILITY

MODES 1, 2, and 3.

ACTIONS

~-------~----NOTES------------------------------------

1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME

A.I~~

ONG- oe Mot:E' i:=l)t..IC.T I ok>S.

eEat.J1"8:>

IAJOR;;t!A8LE.

fAI.

3'.l<a*02 Restore required 3.3.8-07

.....--p.ow.M..1-UDo..u to OPERABLE B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months Palisades Nuclear Plant 3.3.8-1 Amendment No. 01/20/98

Alternate Shutdown System 3.3.8.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.1 Perfonn CHANNEL FUNCTIONAL TEST of the Once within Source Range Neutron Flux Function. 7 days prior to each reactor startup SR 3.3.8.2 Verify each required control circuit and* 18 months transfer switch is capable of perfonning the intended function.

E-D Perfonn CHANNEL CALIBRATION for each 18 months required instrumentation channel.*

2. t.lEOll!ON D'E:TEc..1"'ottS A.eE: e<<L\.lo&l>

RoM ~e- CAAtJ~E:.L c.ALIBEAT'tOA.J .

Palisades Nuclear Plant 3.3.8-2 Amendment No. 01/20/98

Alternate Shutdown System 3.3.8

~il Table 3.3.8-1 (page 1 of 1) !.;,e,-oz;>

Alternate Shutdown System Instrumentation and Controls..----~ I FUNCTION/INSTRUMENT OR CONTROL PARAMETER l~~~*~~r~

1. Source Range Neutron Flux 1

2. Pressurizer Pressure 1

3. Pressurizer Level 1

4. Primary Coolant System (PCS) #1 Hot Leg Temperature - 1.._

5. PCS #2 Hot Leg Temperature . -- - 1

6. PCS #1 Cold Leg Temperature .t

7. PCS #2 Cold Leg Temperature 1

8. Steam Generator {SG) A Pressure 1

9. SG B Pressure 1

10. SG A Wide Range Level 1

11. SG B Wide Range Level 1

12. Safety Injection Refueling Water (SIRW) Tank Level

13. Auxiliary Feedwater (AFW) Flow Indication to SG A

14. AFW Flow Indication to SG B

15. AFW Low Suction Pressure Alarm (P-88) 1 1

1 1

16. AFW Pump P-88 Steam Supply Valve Control 1

17. AFW Flow Control to SG A 1

18. AFW Flow Control to SG B L

Palisades Nuclear Plant 3.3.8-3 Amendment No. 01/20/98

Neutron Flux Monitoring Channels 3.3.9 3.3 INSTRUMENTATION 3.3.9 Neutron Flux Monitoring Channels LCO 3.3.9 Two channels of neutron flux monitori~g instrumentation shall be OPERABLE.

APP LI CAB IL ITY:

ACTIONS CONDITION REQUIRED ACTION COMP~ETION TIME

A. One or more required channel(s) inoperable.

A.1 AND Suspend all operations involving positive reactivity additions.

Inunediately A.2 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Once per ~~

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3.~.9*05 thereafter

Palisades Nuclear Plant 3.3.9-1 Amendment No. 01/20/98

l Neutron Flux Monitoring Channels 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.1 Perfonn CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.9.2~rfonn CHANNEL CALIBRATION. 18 months

/

AJEureD~ DE::.TE~roes Aiee- e::~~L.UDEP "F-~oH THE- CJ1AtJ~L CAu5s<Ar1okl .

Palisades Nuclear Plant 3.3.9-2 Amendment No. 01/20/98

3.3 INSTRUMENTATION ESRV Instrumentation 3.3.10 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Ins~rumentation LCO 3.3.10 Two channels of ESRV Instrumentation shall be OPERABLE.

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

ACTIONS

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

Separate Con.dition entry is al lowed for each channel.

CONDITION REQUIRED ACT ION COMPLETION TIME A. One or more channels A. l Initiate action to Immediately inoperable . isolate the . -

associated ESRV System.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.10.1 Perfonn a CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.10.2 Perfonn a CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.10.3 Perf.onn a CHANNEL CALIBRATION. 18 months Verify high radiation setpoint on each ESRV Instrumentation radiation monitoring channel is ~ 2.2E+5 cpm

Palisades Nuclear Plant 3.3.10-1 Amendment No. 01/20/98

Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perfonn CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I

Perfonn CHANNEL CALIBRATION. 18 months

~~~~~~- NOTE::

---.....& >J.E-U\~~ !)"E":T"e-C..Tt)2'$ ...~ F.:r)(t:.L..LJbEC>

,,:::...eOM ~ti:" c~~~EL CAL\ 6~\\Q)J.,

Palisades Nuclear Plant 3.9.2-2 Amendment No. 01/20/98

RPS. Instrumentation B 3.3.1

B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protective System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a reactor trip to protect against violating the eere specified acceptable fuel design limits and breaching the reactor coolant pressure boundary during Anticipated Operational Occurrences (AOOs). By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.

The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. :

The LSSS, defined in this Specification as the.Allowable Values,, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).

During AOOs, which are those events expected to.occur one or more times during the plant life, the acceptable limits are:

The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling;

Fuel centerline melting shall not occur; and

The Primary Coolant System (PCS) pressure SL of 2750 psia shall not be exceeded.

Maintaining the parameters within the above values ensures that the offsite dose will be within the 10 CFR 50 (Ref. 1) and 10 CFR 100 (Ref. ,2J criteria during AOOs.

Accidents are events that are analyzed even though they are not expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall be maintained within an acceptable fraction of 10 CFR 100 (Ref. 2) limits. Different accident categories allow a different fraction of ,these limits based on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.

Palisades. Nuclear Plant .B 3.3.1-1 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES BACKGROUND The RPS is segmented into four interconnected modules.

(continued) These modules are:

Measurement channels (or pressure switches);

RPSBistable trip units;

Matrix Logic; and

Trip Initiation Logic.

This LCO addresses measurement channels and RPSbistable trip units. It also addresses the automatic bypass removal feature for those trips with Zero Power ModeoperatiRg bypasses. The RPS Logic and Trip Initiation Logic are addressed in LCO 3.3.2, Reactor Protective System (RPS) 11 Logic and Trip Initiation, The role of the measurement 11 channels, RPS aRd bistable trip units, and RPS Bypasses is discussed below.

Measurement Channels Measurement channels, consisting of pressure switches, field transmitters, or process sensors and associated instrumentation, provide a measurable electronic signal based upon the physical characteristics of the parameter being measured. Pressure switches provide a coRtact closure or opeRiRg wheR the measured parameter is Rot HithiR its limit.

The excore Ruclear iRstrumeRtatioR (wide raRge aRd power raRge) aRd the thermal margiR moRitors are coRsidered compoReRts iR the measuremeRt chaRRels.

The 'rt'i de raRge Nuclear IRstrumeRts (N Is) provide a Iii gh Startup Rate trip. There are oRly two wide raRge NI chaRRels. The wide raRge chaRRel sigRal processiRg electroRics are physically mouRted iR RPS cabiRet chaRRels C (NI 01/03) aRd 0 (NI 02/04). Separate bistable trip uRits mouRted withiR the ChaRRel C wide raRge chaRRel drawer supply lligh Startup Rate trip sigRals to RPS chaRRels A aRd C. Separate bistable trip uRits mouRted withiR the ChaRRel D wide raRge d1aRRel drawer pro*vi de IIi gh Startup Rate trip sigRals to RPS chaRRels B aRd D.

Palisades Nuclear Plant B 3.3.1-2 01/20/98 05/30/99

RPS Instrumentation

BASES BACKGROUND Measurement Channels (continued)

B 3.3.1 Two RPS trips use a power level designated as Q Power as an input. Q Power is the higher of NI power from the power range NI drawer and primary calorimetric power (t:.T power) based on PCS hot leg and cold leg temperatures. Trips using Q Power as an input are the Vari ahl e Iii gh Power Trip ('t'llPT) anel the Thermal Margi n/Lmt' Pressure (TM/LP) trips, both of which employ a thermal margin monitor for trip generation.

The thermal margin monitors proviele the complex signal processing necessary to calculate the TM/LP trip setpoint, VllPT trip setpoint anel trip comparison, and Q Power calculation.

The excore power range Nls (NI 005 through NI 008) are mounted in the RPS cabinet, with one channel of each in each of the four RPS hays.

With the exception of Hi Startup Ratethe wiele range Nis, which employs two instrument channels. and Loss of Load, which employs a single pressure sensor, four identical measurement channels with electrical and physical separation are provided for each parameter used in the direct generation of trip signals. These are designated channels A through D. Some measurement channels provide input to more than one or more RPS bistable trip unit within the same RPS channel. In addition, some measurement channels may also be used as inputs to Engineered Safety Features (ESF) bistables, and most provide indication in the control room.

In the case of Hi Startup Rate and Loss of Load, where fewer than four sensor channels are employed, the reactor trips provided are not relied upon by the plant safety analyses.

The sensor channels do however, provide trip input signals to all four RPS channels.

When a channel monitoring a parameter exceeds a predetennined setpoint, indicating an abnonnalunsafe condition, the bistable monitoring the parameter in that channel will trip. Tripping two or more channels of bistable trip units monitoring the same parameter de-energizes Matrix Logic, (addressed by LCO 3.3.2) which in turn de-energizes the Trip Initiation Logic. This causes all four DC clutch power supplies to de-energize, interrupting power to*the control rod drive mechanism ,

clutches, allowing the full length control rods to insert

into the core .

Palisades Nuclear Plant B 3.3.1-3 01/20/98 05/30/99

RPS Instrumentation

BASES BACKGROUND Measurement Channels (continued)

B 3.3.1 For those trips relied upon in the safety analyses, tfhree of the four measurement and trip unit channels are necessary to meet the redundancy and testability of GDC 21 in 10 CFR 50, Appendix A (Ref. 1). The fourth channel provides additional flexibility by allowing one channel to be removed from service (trip channel bypass) for maintenance or testing while still maintaining a minimum two-out-of-three logic.

Si nee no single failure wi 11 prevent a protective system actuation, this arrangement meets the requirements of IEEE Standard 279-1971 (Ref. 3).

In the ease ef wide range pewer and Less ef Lead, where fewer than.fettr senser channels are empleyed, the reaeter trips previded are net reqttired. by the plant safety analyses. As stteh, they need net meet the abeve criteria.

In these eases, hewever, the senser channels de however, previde trip inpttt s~gnals te all fettr RPS channels .

There are twe types ef trip ttnits, a bistable trip ttnit and an. attxiliary trip ttnit. A bistable trip ttnit eempares a precess signal tea setpeint. An attxiliary trip ttnit ttses a digital inpttt (eentaets epen er elesed). Each RPS channel has fettr attxi~iary trip ttnits and seven bistable trip ttnits.

Most of the RPS trips are generated by comparing a single measurement to a fixed bistable setpoint. Two trip Certain Functions, Variable High Power Trip and Thermal Margin Low Pressure Trip, hewever, make use of more than one measurement to provide a trip.

The fellowing trips ttse mttltiple measttrement channel inpttts.

The required RPS Trip Functions utilize the following input instrumentation:

Variable High Power Trip (VHPT)

The VHPT uses Q Power as its ~input. Q Power is the hi~her of NI power from the power range NI drawer and primary calorimetric power (~T power) based on PC~

hot leg and cold leg temperatures. The measurement channels associated with the VHPT are the power range excore channels; and the PCS hot and cold leg -

temperature channels .

Palisades Nuclear Plant B 3.3.1-4 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels

Variable High Power Trip (VHPT) (continued)

The Thermal Margin Monitors provide the complex signal processing necessary to calculate the TM/LP trip setpoint, VHPT trip setpoint and trip comparison, and Q Power calculation. Q Pewer is the higher ef NI pewer and 6T pewer. It has a trip On power decreases the VHPT setpoint -tfla-t- tracks power levels downward so that it is always within a fixed increment above current power, subject to a minimum value.

On power increases, the trip setpoint remains fixed unless manually reset, at which point it increases to the new setpoint, a fixed increment above Q Power at the time of reset, subject to a maximum value. Thus, during power escalation, the trip setpoint must be repeatedly reset to avoid a reactor trip.

High Startup Rate Trip I

The High Startup Rate trip uses the wide range nuclear instruments (Nls) to provide an input signal. There are only two wide range NI channels. The wide range channel signal processing electronics are physically mounted in RPS cabinet channels C (NI-1/3) and D (NI-2/4). Separate bistable trip units mounted within I

I I

the NI-1/3 wide range channel drawer supply High I Startup Rate trip signals to RPS channels A and C. I Separate bistable trip units mounted within the NI-2/4 I wide range channel drawer provide High Startup Rate I trip signals to RPS channels B and D. I

Low Primary Coolant Flow Trip The Low Primary Coolant Flow Trip utilizes 16 flow measurement channels which monitor the differential pressure acros.s the primary side of the steam generators. Each RPS channel, A, B, C, and D, receives a signal which is the sum of four differential pressure signals. This totalized si~nal is compared with a setpoint in the RPS Low Flow bistable trip unit for that RPS channel .

Palisades Nuclear Plant B 3.3.1-5 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)

Low Steam Generator Level Trips I I

There are two separate Low Steam Generator Level I trips, one for each steam generator. Each Low Steam I Generator Pressure trip monitors four level I measurement channels for the associated steam I generator, one for each RPS channel. I.

I I

Low Steam Generator Pressure Trips I There are also two separate Low Steam Generator Pressure trips, one for each steam generator. Each Low Steam Generator Pressure trip monitors four pressure measurement channels for the associated steam generator, one for each RPS channel.

High Pressurizer Pressure Trip

The High Pressurizer Pressure Trip monitors four pressurizer pressure channels, one for each RPS channel.

Thermal Margin Low Pressure (TM/LP) Trip The TM/LP Trip utilizes bistable trip units. Each of these bistable trip units receives a calculated trip setpoint from the Thermal Margin Monitor (TMM) and compares it to the measured pressurizer pressure signal. The TM/LP setpoint is based on Q power (the higher of NI power from the power range NI drawer, or

~T power, based on PCS hot leg and cold leg temperatures) pressurizer pressure, PCS cold leg temperature, and Axial Shape Index. The TMM provide the complex signal processing necessary to calculate the TM/LP trip setpoint, TM/LP trip comparison signal, and Q Power .

Palisades Nuclear Plant B 3.3.1-6 01/20/98 05/30/99

RPS Instrumentation

BASES BACKGROUND Measurement Channels (continued)

B 3.3.1

nermal Marai r=1!Lo*w Pressure (TM/LP) p I Q Power is or=1ly or=1e of several i flputs to the TP1/LP trip. Other iflputs ir=1cluee iflterflal ASI ar=1e cole leg temperature basee Ofl the higher of two cole leg resistar=1ce temperature eetectors. The TM/LP trip setpoiflt i~ a complex fuflctiofl of these iflputs afle represer=1ts a mir=1imum acceptable PCS pressure to be comparee to actual PCS pressure ifl the TM/LP trip t:ttrt-t-;.

Loss of Load Trip The Loss of Load Trip is actuated by turbine auxiliary relays 305L and 305R. Relay 305L provides input to RPS channels A and C; 305R to channels B and D.

Relays 305L and 305R are energized on a turbine trip.

Their inputs are the same as the inputs to the turbine soleniod trip valve, 20ET .

If a turbine trip is generated by loss of auto stop oil pressure, auto stop oil pressure switch 63/AST-2 will actuate relays 305L and 305R and generate a reactor trip. If a turbine trip is generated by an input to the soleniod trip valve, relays 305L and

1 I

I I

I 305R, which are wired in parallel, will also be I actuated and will generate a reactor trip. I I

Containment High Pressure Trip The Containment High Pressure Trip is actuated by four pressure switches, one for each RPS channel.

Zero Power Mode Bypass Automatic Removal The Zero Power Bypass allows manually bypassing (i.e.

disabling) four reactor trip functions, Low PCS Flow, Low SG A Pressure, Low SG B Pressure, and TM/LP (low PCS pressure), when reactor power (as indicated by the wide range nuclear instrument channels) is below 10-4%. This bypass,ing is necessary to allow RPS testing and control rod drive mechanism testing when the reactor is shutdown and plant conditions would cause a reactor trip to be present.

Palisades Nuclear Plant B 3.3.1-7 01/20/98 05/30/99

RPS Instrumentation

BASES BACKGROUND Measurement Channels I

(continued)

B 3.3.1

Zero Power Mode Bypass Automatic Removal (continued)

The Zero Power Mode Bypass removal interlock uses the wide range nuclear instruments (Nis) as measurement channels. There are only two wide range NI channels.

Separate bistables are provided to actuate the bypass removal for each RPS channel. Bistables in the NI-1/3 channel provide the bypass removal function for RPS channels A and C; bistables in the NI-2/4 channel for RPS channels B and D.

Several measurement instrument channels provide more than one required function. Those sensors shared for RPS and ESF functions are identified in Table B 3.3.1-1. That table provides a listing of those shared channels and the Specifications which they affect.

RPSBistable Trip Units Two types of RPS trip units are used in the RPS cabinets; bistable trip units and auxiliary trip units:

A bistable trip unit recieves a me~sured process signal from its instrument channel and compares it to a setpoint; the trip untt actuats three relays, with contacts in the Matrix Logic channels, when the measured signal is less conservative than the setpoint. They also provide local trip indication and remote annunciation.

An auxiliary trip unit recieves a digital input (contacts open or closed);the trip unit actuats three relays, with contacts in the Matrix Logic channels, when the digital input is recieved. They also provide local trip indication and remote annunciation.

Each RPS channel has four auxiliary trip units and seven bistable trip units.

Bistable trip Ufiits, mouHted ifi the RPS cabiHet, receive aft aHalog iHput from the measuremefit chaHHels, compare the aHalog iHput to trip setpoiHts, aHd provide coHtact output to the ~'atrix Logic. They also provide local trip iHdicatiofi afid remote'afifiUficiatiofi .

Palisades Nuclear. Plant B 3.3.1-8 01/20/98 05/30/99

RP~ Instrumentation

BASES BACKGROUND Measurement Channels B 3.3.1 RPSBistable Trip Units (continued)

The contacts from these bistable trip unit relays are arranged into six coincidence matrices, comprising the Matrix Logic. If bistable trip units monitoring the same parameter in at least two channels trip, the Matrix Logic will generate a reactor trip (two-out-of-four logic).

Four of the RPS measurement channels provide contact outputs to the RPS, so the comparison of an analog input to a trip setpoint is not necessary. In these cases, the bistable trip unit is replaced with an auxiliary trip unit. The

auxiliary trip units provide contact multiplication so the single input contact opening can provide multiple contact outputs to the coincidence logic as well as trip indication and annunciation.

Trips employing auxiliary trip units include the VHPT, which receives contact inputs from the Thermal Margin Monitors; the High Startup Rate trip which employs contact inputs from bistables mounted in the two wide range drawers; the Loss of Load Trip which receives contact inputs from one of two 1

auxiliary relays which are operated by a single switch sensing turbine Electre llydraulic Cefltrel (EllC) auto stop oil pressure; and the Containment High Pressure (CHP) trip, which employs*containment pressure switch contacts.

There are four RPSchaflflels ef bistable trip units, designated as channels A through D, each channel having eleven trip units, one for each RPS Function, efle fer each measuremeflt chaflflel. Bistable Trip unit output relays de-energize when a trip occurs.

All RPS Trip Functionst-r-1-f}s-, with the exception of the Loss of Load and CHP trips, generate a pretrip alarm as the trip setpoint is approached .

Palisades Nuclear Plant B 3.3.1-9 01/20/98 05/30/99

RPS Instrumentation

BASES BACKGROUND RPSBistable Trip Units (continued)

B 3.3.1 The trip.setpoiRts used iR the bistable trip uRits are based OR the aRalytical limits stated iR RefereRce 4. The selection of these trip setpoiRts is such that adequate protection is provided wheR all seRsor aRd processing time delays are takeR iRto account. To allow for calibratioR toleraRces, iRstrumentatioR uRcertaiRties, aRd iRstrumeRt drift, Allowable Values specified iR Table 3.3.1 1, iR the*

accompanyiRg LCO, are coRservatively adjusted with respect to the analytical limits. The methodology used to validate the trip setpeiRts is provided in plant documents. The nemiRal trip setpoiRt entered iRto the bistable is Rormally still mere coRservative thaR that specified by the Allowable Value. to accouRt for chaRges iR raRdom measurement errors.

A chaRRel is iRoperable if its actual setpoiRt is Rot within its required Allowable Value.

The Allowable Values are specified for each.safeiy related*

RPS trip Function which is credited in the safety analysis.

Nominal trip setpoints are specified in *the plant

procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required.* The me'thedology used to determine the nominal trip setpoints is also provided in plant documents. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value

specified is more conservative than the analytical limit determined* in the safety analysis in order to account for uncertainties appropriate to the trip Function. These uncertainties are addressed as described.iri plant documents.

A channel is inoperable if its actual s~tpoint is not within its Allowable Value.

Setpoints in accordance with the Allowable Value will ensure that Sls of Chapter 2.0 are not violated during AOOs and the consequences of DBAs will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or OBA and the equipment functions as designed.

I Note that in the accompanying LCO 3.3.l, the Allowable Values of Table 3.3.1-1 are the LSSS.

General Palisades Nuclear Plant B 3.3.1-10 01/20/98 05/30/99

RPS Instrumentation

BASES BACKGROUND (continued)

Reactor Protective System Bypasses B 3.3.1 I

Three different types of trip bypass are utilized in the I RPS, Operating Bypass, Zero Power Mode Bypass, and Trip I Channel Bypass. The Operating Bypass or Zero Power Mode I Bypass prevent the actuation of a trip unit or auxiliary I trip unit; the Trip Channel Bypass prevents the trip unit I outpu~ from affecting the Logic Matrix. A channel which is I bypassed, other than as allowed by the Table 3.3.1-1 I footnotes, cannot perform its specified safety function and I must be considered to be inoperable. I I

Operating Bypasses I I

The Operating Bypasses are initiated and removed I automatically during startup and shutdown as *power level_

changes. An Operating Bypass prevents the associated RPS auxiliary trip unit from receiving a trip signal from the associated measurement channel. With the bypass in place~

neitherthe pre-trip alarm nor the trip will actuate if the measured parameter exceeds the set point. An annunciator is

provided for each Operating Bypass. The RPS trips with Operating Bypasses are:

a. High Startup Rate Trip bypass. The High Startup Rate trip is automatically bypassed when the associated wide range channel indicates below lE-4% RTP, and when the associated power range excore channel indicates above 13% RTP. These bypasses are autom~tically removed between lE-4% RTP and 13% RTP.

b. Loss of Load bypass. The Loss of Load trip is automatically bypassed when the associated power range excore channel indicates below 17% RTP. The bypass is automatically removed when the channel indicates above the set point. The same power range excore channel bistable is used to bypass the High Startup Rate trip and the Loss of Load trip for that RPS channel.

Each wide range channel contains two bistables set at lE-4%

RTP, one bistable unit for each associated RPS channel.

Each of the two wide range channel affects the Operating Bypasses for two RPS channels; wide range channel NI-1/3 for RPS channels A and C, wide range channel NI-2/4 for RPS channels B and D. Each of the four power range excore channel affects the Operating Bypasses for the associated RPS channel. The power range excore channel bistables

associated with the Operating Bypasses are set at a nominal 15%, and are required to actuate between 13% RTP and 17% RTP.

Palisades Nuclear Plant B 3.3.1-11 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES BACKGROUND Zero Power Mode (ZPM) Bypass I (continued) I The ZPM Bypass is used when the plant is shut down and it is I desired to raise the control rods for control rod drop I testing with PCS flow, pressure or temperature too low for I the' RPS trips to be reset. ZPM bypasses may be manually I initiated and removed when wide range power is below lE-4% I RTP, and are automatically removed if the associated wide I.

range NI indicated power exceeds lE-4% RTP. A ZPM bypass prevents the RPS trip unit from actuating if the measured parameter exceeds the set point. Operation of the pretrip alarm is unaffected by the zero power mode bypass. An annunciator indicates the p~esence of any ZPM bypass. The

. RPS trips with ZPM bypasses are: *

a. Low Primary Cool ant System Fl ow.

b. Low Steam Generator Pressure.

c. Thermal Margin/Low Pressure .

. The wide range NI channels provide contact closure permissive signals when indicated power is below .lE-4% RTP.

The ZPM bypasses may then be manually initiated or removed by actuation of key-1 ock switches. One key-1 ock swi.tch located on each RPS cabinet controls the ZPM Bypass for th~ .*

associated RPS trip channels. The bypass is automatically removed if the associated wide range NI indicated power exceeds lE-4% RTP. The same wide range NI channel bistables that provide the ZPM Bypass permissive and removal signals also provide the high startup rate trip Operating Bypass actuation and removal.

Trip Channel Bypass A Trip Channel Bypass is used when it is desired to physically remove an individual trip unit from the system, or when calibration or servicing of a trip channel could cause an inadvertent trip. A trip Channel Bypass may be manually initiated or removed at any time by actuation of a key-lock switch. A Trip Channel Bypass preveAts the trip unit output from effecting the RPS logic matrix. A light above the bypass switch** indicates that the trip channel has been bypassed. Each RPS trip unit has an associated'trip channel bypass:

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RPS Instrumentation

BASES BACKGROUND Trip Channel Bypass (contineud)

B 3.3.1 The key-lock trip channel bypass switch is located above each trip unit. The key cannot be removed when in the bypass position. Only one key for each trip parameter is provided, therefore the operator can bypass only one channel of a given parameter at a time. During the bypass condition, system logic changes from two-out-of-four to two-out-of-three channels required for trip.

In addition te the trip channel bypasses, there are alse eperating bypasses on select RPS trips. Some ef these operating bypasses are enabled manually, others autematically, iH all four channels when plant conditions do Hot warrant the specific trip protection. All eperating bypasses are automatically removed when enabling bypass conditions are RO longer satisfied. Trips with operating bypasses include the lligh Startup Rate, Low PCS. Flow, Low Steam Generater (SG) Pressure, TM/LP, and Loss of Load. The Less of Lead trip and lligh Startup Rate trip operatiRg bypasses are autematically .enabled and disabled .

Several instrument channels provide mere than one.required function. Table B 3.3.1 1 provides a listing of these Fhannels and the Specifications which they affect.

APPLICABLE Each of the analyzed accidents and transients can be SAFETY ANALYSES detected by* one or more RPS Functions. The accident analysis contained in Reference 4 takes credit for most RPS trip Functions. The High Startup Rate and Loss of Load Functions, which are not specifically credited in the accident analysis are part of the NRC approved licensing basis for the plant. The High Startup Rate and Loss of Load trips are purely equipment protective, and their use minimizes the potential for equipment damage.

The specific safety analyses applicable to each protective Function ~re identified below.

1. Variable High Power Trip (VHPT)

The VHPT provides reactor core protection against positive reactivity excursions.

The safety analysis assumes that this trip is OPERABLE to tenninate excessive positive reactivity insertions during power operation and while shut down.

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RPS Instrumentation B 3.3.1 BASES APPLICABLE 2. High Startup Rate Trip SAFETY ANALYSES (continued) There are no safety analyses which take credit for functioning of the High Startup Rate Trip. The High Startup Rate trip is used to trip the reactor when excore wide range power indicates an excessive rate of change. The High Startup Rate trip minimizes transients for events such as a continuous control rod withdrawal or a boron dilution event from low power levels. The trip may be operationally bypassed when THERMAL POWER is < lE-4% RTP, when poor counting statistics may lead to erroneous indication. It may also be operationally bypassed at > 13% RTP, where moderator temperature coefficient and fuel temperature coefficient make high rate of change of power unlikely. IR MODES 3, 4, aftd 5 wheft RO more thaft Ofte cofttrol rod is capable of beiRg withdrawR or the PCS boroft COftCefttratioR is at the REFUELING BORON CONCENTRATION, the lligh Startup Rate trip is Rot required to be OPERABLE, hbwever, the iRdicatioR aftd alarm FuRctioRs of both wide raRge chaRRels are re~uired by LCD 3.3.9, "NeutroR Flux MoRitoriRg ChaMels," to be OPERABLE. LCO 3.3.9 eftsures the wide raRge chaRRels are available to detect aRd alert the operator to a boroR dilutioR eveRt.

There are only two wide range drawers, with each supplying contact input to auxiliary trip units in two RPS channels.

3. Low Primary Coolant System Flow Trip The Low PCS Flow trip provides DNB protection during events which suddenly reduce the PCS flow rate during power operation, such as loss of power to, or seizure of, a primary coolant pump.

Flow in each of the four PCS loops is determined from pressure drop from inlet to outlet of the SGs. The total PCS flow is determined, for the RPS flow channels, by summing the loop pressure drops across the SGs and correlating this pressure sum with the sum of SG differential pressures which exist at 100% flow (four pump operation at full power Tave). Full PCS flow is that flow which exists at RTP, at full power Tave' with four pumps operating .

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RPS Instrumentation B 3.3.1 BASES APPLICABLE 4, 5. Low Steam Generator Level Trip SAFETY ANALYSES (continued) The Low Steam Generator Level trips are provided to trip the reactor in the event of excessive steam demand (to prevent overcooling the PCS) and loss of feedwater events (to prevent overpressurization of the PCS}. 1 The Allowable Value assures that there will be sufficient water inventory in the SG at the time of trip to allow a safe and orderly plant shutdown and to prevent SG dryout assuming minimum AFW capacity.

Each SG level is sensed by measuring the differential pressure between the top and bottom of in the upper portion of the. downcomer annulus in the SG. These trips share four level sensing channels on each SG with the AFW actuation signal.

- 6, 7. Low Steam Generator Pressure Trip The Low Steam Generator Pressure trip provides protection against an excessive rate of heat extraction from the steam generators, which would result in a rapid uncontrolled cooldown of the PCS.

This trip provides a mitigation function in the event of an MSLB.

The Low SG Pressure channels are shared with the Low SG Pressure signals which isolate the steam and feedwater lines.

8. High Pressurizer Pressure_Trip The High Pressurizer Pressure trip, in conjunction with pressurizer safety valves and Main Steam Safety Valves (MSSVs), provides protection against overpressure conditions in the PCS when at operating temperature. The safety analyses assume the High Pressurizer Pressure trip is OPERABLE during accidents and transients which suddenly reduce PCS cooling (e.g., Loss of Load, Main Steam Isolation Valve (MSIV) closure, etc.) *or which suddenly increase reactor power (e.g., rod ejection accident) .

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RPS Instrumentation B 3.3.1 BASES APPLICABLE 8. High Pressurizer Pressure Trip (continued)

SAFETY ANALYSES The High Pressurizer Pressure trip shares four safety grade instrument channels with the TM/LP trip, Anticipated Transient Without Scram (ATWS) and PORV circuits, and the Pressurizer Low Pressure Safety Injection Signal.

9. Thennal Margin/Low Pressure (TM/LP) Trip The TM/LP trip is provided.to prevent reactor operation when the DNBR is insufficient. The TM/LP trip protects against slow reactivity or temperature increases, and against pressure decreases.

The trip is initiated whenever the PCS pressure signal drops below a minimum value (P~") or a computed value (P~r) as described below, whichever is higher.

The TM/LP trip uses Q Power, ASI, pressurizer pressure, and cold leg temperature (Tc) as inputs.

Q Power is the higher of core THERMAL POWER (~T Power) or nuclear power. The ~T power uses hot leg and cold leg RTDs as inputs. Nuclear power uses the power range excore channelsRuclear iRstrumeRts as inputs.

Both the ~T and excore power signals have provisions for calibration by calorimetric calculations.

The ASI is calculated from the upper and lower power range excore detector signals, as explained in Section 1.1, "Definitions." The signal is corrected for the difference between the flux at the core periphery and the flux at the detectors.

The Tc value is the higher of the two cold leg signals.

The Low Pressurizer Pressure trip limit (P var) is calculated using the equations given in Table 3.3.1-2 .

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. I RPS Instrumentation B 3.3.1 BASES APPLICABLE 9. Thennal Margin/Low Pressure (TM/LP) Tri.Q (continued)

SAFETY ANALYSES The .calculated limit (Pvar) is then compared to a fixed Low Pressurizer Pressure.trip limit (P~n). The auctioneered highest ofef' these signals becomes the trip limit (Ptrip). Ptrip is. compared to the measured PCS pressure and a trip signal is generated when the measured pressure for that channel is less than or equal to Ptrip* A pre-trip alann is also generated when P is less than or equal to the pre-trip setting, Ptrip + ti.P

The TM/LP trip setpoint is a complex function of these inputs and represents a minimum acceptable PCS pressure for the existing temperature and power conditions. It is compared to actual PCS pressure in the TM/LP trip unit.

10

Loss of Load Trip

There are no safety analyses which take credit for functioning of the Loss. of Load Trip.

The Loss of Load trip is provided to prevent lifting the pressurizer and main steam safety valves in the event of a turbine generator trip while at power. The trip is equipment protective. The safety analyses do not assume that this trip functions during any accident or transient. The Loss of Load trip uses a sin[le pressure switch in the turbine auto stop oil circuit to sense a turbine trip for input to all four RPS auxiliary trip units.

11. Containment High Pressure Trip I The Containment High Pressure trip provides a reactor trip in the event of a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). The Containment High Pressure trip shares sensors with the Containment High Pressure sensing logic for Safety Injection, Containment Isolation, and Containment Spray. Each of these sensors has a single bellows which actuates two mi croswi tches. ,one mi croswi tch on each of four sensors provides an input to the RPS .

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RPS Instrumentation B 3.3.1 BASES APPLICABLE Bypasses SAFETY ANALYSES (continued) 12. Zero Power Mode Bypass Removal The only RPS bypass considered in the safety analyses is the Zero Power Mode (ZPM) Bypass. The ZPM Bypass is used when the plant is shut down and it is desired to raise the control rods for control rod drop testing with PCS flow or temperature too low for the RPS Low PCS Flow, Low SG Pressure, or Thermal Margin/Low Pressure trips to be reset. ZPM bypasses are automatically removed if the wide range NI indicated power exceeds lE-4% RTP.

The safety analyses take credit for automatic removal of the ZPM Bypass if reactor criticality due to a Continuous Control Rod Bank Withdrawal should occur with the effected trips byp_assed and PCS flow, pressure, or temperature below the values at which the RPS could be reset. The ZPM Bypass would effectively be removed when the first wide range NI channel indication reached lE-4% RTP. With the ZPM Bypass for two RPS channels removed, the RPS would trip on one of the un-bypassed trips. This would prevent the reactor reaching an excessive power level.

If a reactor criticality due to a Continuous* Control Rod Bank Withdrawal should occur when PCS flow, steam generator pressure, and PCS pressure (TM/LP) were above their trip set points, a trip would terminate the event when power increased to the minimum setting (nominally 30%) of the Variable High Power Trip. In this case, the monitored parameters are at or near their normal operational values, and a trip initiated at 30% RTP provides adequate protection.

The RPS design also includes automatic removal of the Operating Bypasses for the High Startup Rate and Loss of Load trips. The safety analyses do not assume functioning of either these trips or the automatic removal of their bypasses.

The bypasses aRd their settiRgs are addressed iR feetRetes te Table 3.3.1 1.

The RPS eperatiRg bypasses are .

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RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES

12. Zero Power Mode Bypass Removal (continued)

a. Iii gh Start1:1p Rate bypass. The Iii gh Start1:1p Rate trip may be a1:1tematieally bypassed at E lE 4% RTP, as sensed by wide range NI bistables, and at ~ 13% RTP by the pewer range NI bistable.

b. Loss of Load bypass. The Loss of Load trip may be a1:1tomatieally bypassed when at E 17% RTP as sensed by the power range NI bistable. The bypass is a1:1tomatically removed by this bistable above the setpoint. This same bistable is 1:1sed to bypass the lligh Start1:1p Rate trip.

One other eperatienal bypass is provided in the RPS design, b1:1t it eannet be 1:1sed when the ~PS is req1:1ired te be OPERABLE. The Zere Power Mode Bypass (ZPMB) is man1:1ally aet1:1ated. Man1:1al aet1:1atien is enabled when both wide range NI channels are below lE 4~ RTP, and the bypass is a1:1temati tally remeved whefl either chaflflel is above that setpoint. The ZPMB disables the TM/LP, Low SG Press1:1re, afld Lew PCS Flow trips. This operatioflal bypass allows eontrel rod testing whefl PCS press1:1re, flow, or temperat1:1re.is toe low to allow resetting the trips.

The wide range fl1:1x level ifldieatiofl aet1:1ates bistables wl=li ch act1:1ate the permi sshe si gflal for the ZPMB (for the TM/LP, Low PCS Flew, and Low SG Press1:1re trips), and bypass the Iii gh Start1:1p Rate trip below lE 4°~ RTP. Wide range ehanflel NI l/3A provides the bypass permissive'fer RPS ehanflels A afld C, NI 2/4A for chaflnels B and D. A separate bistable trip 1:1nit is previded for each RPS chaflflel.

The same bistables _that provide the ZPMB permissi*ve also a1:1temati cal ly bypass the Iii gh Start1:1p Rate trips bel ew the setpoi nt afld eflabl e them ab eve. Whefl at very 1e*,; power levels, the f11:1clear iflstr1:1me~t signals are flot steady, if the lligh Start1:1p Rate trips were flet bypassed, sp1:1rio1:1s trips eo1:1ld eec1:1r d1:1riflg start1:1p eperatiofls .

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RPS Instrumentation B 3.3.1 BASES APPLICABLE Bvpasses (coRtiRued)

SAFETY ANALYSES Hie lligh Startup Rate trip is automatically Bypassed when power range indicated power exc~eds a nominal 15% RTP.

Allowing for hysteresis, this Bypass may Be as low as 13% RTP. The trip is not useful aBove that power level since reactivity insertions at power would induce an immediate change in power level and eventually Be terminated by the \'llPT without attaining any significant startup rate.

This bypass is automatically removed when the associated power range indication decreases below the Bistable setpoint.

Power range NI OS provides the bistable for RPS channel A, NI 06 for channel B, NI 07 for channel C, and NI 08 for channel D. These same power range bistable amplifiers also bypass the Loss of Load trip below the setpoint.

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

LCO The LCO requires all instrumentation performing an RPS Function to be OPERABLE. Failure of the trip unit (including its output relays), any required portion of the associated instrument channel, or both, renders the affected channel(s) inoperable and reduces the reliability of the affected Functions. Failure of an-tfte automatic ZPM (operating) bypass removal channel may also impact the associated instrument channel(s) and reduce the reliability of the affected Functions. The specific criteria for determining channel OPERABILITY differ slightly Between Functions. These criteria are discussed on a Function BY Function Basis Below.

Actions allow Trip Channel Bypassmaintenance (trip channel)

Bypass of individual channels, but the bypassed channel must be considered to be inoperable. Tthe bypass key used to bypass a single channel cannot be simultaneously used to bypass that same parameter in other channels. This interlock prevents operation with more than one channel of the same Function bypassed. The plant is normally restricted to 7 days in a trip channel bypass, or otherwise inoperable condition before either restoring the Function to four channel operation (two-out-of-four logic) or placing

the channel in trip (one-out-of-three logic) .

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RPS Instrumentation B 3.3.1 BASES LCO The Allowable Values are specified for each safety related (continued) RPS trip Function in the LCO. Nominal trip setpoints are

/ specified in the plant procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument leep is performing as required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable, provided that eperatien and testing are eensistent with the assumptions ef the setpeint calculations. Each 1Allowable Value specified is mere conservative than the analytical limit assumed in the safety analysis in order te account for instrument uncertainties appropriate to the trip Function. These uncertainties are addressed as described in plant documents. Neither allowable values nor setpoints are specified for the non safety related RPS Trip Functions, since no safety analysis assumptions would be violated if they are not set at a particular value.

The Allowable Values are specified for each safety related

.RPS trip Function which is credited in the safety analysis

Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure plant param~ters do not exceed the Allowable Valu~ if the .

instrument loop is performing as required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.

Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function. These uncertainties are addressed as described in p1ant documents. Neither A11 ow ab 1e Va 1ues nor set points ar*e specified for the non-safety related RPS Trip Functions, since no safety analysis assumptions would be violated if they are not set at a particular value.

The following Bases for each trip Function identify the above RPS trip Function criteria items that are applicable to establish the trip Function OPERABILITY.

This LCO requires that all four channels of all trip Functions be OPERABLE when in MODES 1 and 2, and in MODES 3, 4, and 5 whenever more than one control rod is capable of being ~ithdrawn and the PCS beren concentration is less than REFUELI~G BORON CONCENTRATION. Exceptions are noted in the individual Function LCO Bases belew .

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BASES LCO 1. Variable Hiah Power Trip (VHPT)

RPS Instrumentation B 3.3.1 (continued)

This LCO requires all four channels of the VHPT Functiori to be OPERABLE.

The Allowable Value is high enough to provide an operating envelope that prevents. unnecessary VHPT trips during normal plant operations. The Allowable Value is low enough for the system to function adequately during reactivity addition events.

The VHPT is designed to limit maximum reactor power to its maximum design and to terminate power excursions initiating at lower powers without power reaching this full power limit. During plant. startup, the VHPT trip setpoint is initially at its minimum value, ~ 30%.

Below 30% RTP, the VHPT setpoint is not required to.

"track" with Q Power, i.e., be adjusted to within 15% RTP. It remains fixed until manually reset, at which point it increases to ~ 15% above existing Q Power .

The maximum allowable setting of the VHPT is 106.5% RTP. Adding to this the possible variation in trip setpoint due to calibration and jnstru~ent error, the maximum actual steady state power at which a trip would be actuated is 115%, which is the value assumed in the safety analysis.

2. High Startup Rate Trip This LCD requires four channels of High Startup Rate Trip Function to be OPERABLE in MODES 1 and 2. The High Startup Rate trip may be bypassed when the wide range NI indicates below lOE-4% or when THERMAL POWER is above 13% RTP. If a High Startup Rate trip is bypassed when power is between these limits, it must be considered to be inoperable.

The High Startup Rate trip serves as a backup to the administratively enforced startup rate limit. The Function is not credited in the accident analyses; therefore, no Allowable Value for the trip or operating bypass Functions is derived from analytical limits and none is ~pecified .

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RPS Instrumentation

BASES LCO 2. High Startup Rate Trip (continued)

B 3.3.1 The four channels of the High Startup Rate trip are derived from two wide range NI signal processing drawers. Thus, a failure in one wide range channel could render two RPS channels inoperable. It is acceptable to continue operation in this condition because the High Startup Rate trip is not credited in any safety analyses.

3. Low Primary Coolant System Flow Trip This LCD requires four channels.of Low PCS Flow Trip Function to be OPERABLE.

This trip is set high enough to maintain fuel integrity during a loss of flow condition. The setting is low enough to allow for normal operating fluctuations from offsite power.

The Low PCS Flow trip setpoint of 95% of full PCS flow insures that the reactor cannot operate when the flow rate is less than 93% of the nominal yalue considering instrument errors. Full PCS flow is that flow which exists at RTP, at full power Tave* with four pumps operating.

4, 5. Low Steam Generator Level Trip Th1s LCD requires four channels of Low Steam Generator Level Trip Function per steam generator to be OPERABLE.

The 25.9% Allowable Value assures that there is an adequate water inventory in the steam generators the heat transfer surface (tubes) is covere~ with water when the reactor is critical and is based upon narrow range instrumentation. The 25.9% indicated level corresponds to the location of the feed ring .

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RPS Instrumentation

BASES LCO (continued) 6, 7. Low Steam Generator Pressure Trip B 3.3~1 This LCO requires four channels of Low Steam Generator Pressure Trip Function per steam generator to be OPERABLE.

The Allowable Value of 500 psia is sufficiently below the full load operating value for steam pressure so as not to interfere with nonnal plant operation. but still high enough to provide the required protection in the event of excessive steam demand. Since excessive steam demand causes the PCS to cool down.

resulting in positive reactivity addition to the core.

a reactor trip is required to offset that effect.

8. High Pressurizer Pressure Trip This LCO requires four channels of High Pressurizer Pressure Trip Function to be OPERABLE.

The Allowable Value is set high enough to allow for pressure increases in the PCS during nonnal operation (i.e .* plant transients) not indicative of an abnonnal condition. The setting is below the lift setpoint of the pressurizer safety valves and low enough to initiate a reactor trip when an abnormal condition is indicated.

9. Thennal Margin/Low Pressure (TM/LP) Trip This LCO requires four channels of TM/LP Trip Function to be OPERABLE.

The TM/LP trip setpoints are derived from the core thennal limits through application of appropriate allowances for measurement uncertainties and processing errors. The allowances specifically account for instrument drift in both power and inlet temperatures. calorimetric power measurement. inlet temperature measurement. and primary system pressure measurement.

Other uncertainties including allowances for assembly power tilt. fuel ,pellet manufacturing tolerances. core flow measurement uncertainty and core bypass flow.

inlet temperature mea~urement time delays. and ASI measurement. are included in the development of the TM/LP trip setpoint used in the accident analysis.

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RPS Instrumentation B 3.3.1 BASES LCO 10. Loss of Load Trio (continued)

The LCO requires four Loss of Load Ttrip Function channels to be OPERABLE in MODE 1 with THERMAL POWER

~ 17% RTP.

The Loss of Load trip may be bypassed or be inoperable with THERMAL POWER < 17% RTP, since it is no longer needed to prevent lifting of the pressurizer safety valves or steam generator safety valves in the event of a Loss of Load. Loss of Load Trip unit must be considered inoperable if it is bypassed when THERMAL POWER is above 17% RTP.

This LCO requires four RPS Loss of Load auxiliary trip units. relays 305L and 305R, and pressure switch 63/AST-2 to be OPERABLE. With those components OPERABLE, a turbine trip will generate a reactor trip.

The LCO does not require the various turbine trips.

themselves, to be OPERABLE .

The Nuclear Steam Supply System and Steam Dump System are capable of accommodating the Loss of Load without requiring the use of the above equipment.

The Loss of Load Trip Function is not credited in the accident analysis; therefore. an Allowable Value for the. trip cannot be derived from analytical limits. and is not specified.

11. Containment High Pressure Trip This LCO requires four channels of Containment High Pressure Trip Function to be OPERABLE.

The Allowable Value is high enough to allow for small pressure increases in containment expected during nonnal operation (i.e .* plant heatup) that are not indicative of an abnonnal condition. The setting is low enough to initiate a reactor trip to prevent containment pressure from exceeding design pressure following a OBA and ensures the reactor is shutdown before initiation of safety injection and containment spray .

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RPS Instrumentation B 3.3.1 BASES LCO 12 ZPM Bvoass I.

(continued)

The LCO requires that four channels of automatic Zero Power Mode (ZPM) Bypass removal instrumentation be OPERABLE. Each channel of automatic ZPM Bypass removal includes a shared wide range NI channel, an actuating bistable in the wide range drawer, and a relay in the associated RPS cabinet. Wide Range NI cha~nel 1/3 is shared between ZPM Bypass removal channels A and C; Wide Range NI channel 2/4, between ZPM Bypass removal channels B and D. An operable bypass removal channel must be capable of automatically removing the capability to bypass the affected RPS trip channels with the ZPM Bypass key switch at the proper setpoint.

Bypasses The LCO on l:lyr:iass r:iermissive removal channels requires that the atJtomati c l:lyr:iass removal feature of all __four or:ierati rig l:lyr:iass channels !:le OPERABLE for each RPS Furiction with ari or:ierating l:lyr:iass in the MODES addressed iri the sr:iecific LCO for each Functiori. All four l:lyr:iass removal chaHHels must !:le OPERABLE to ensure that riorie of the fo~r RPS channels are inadvertently l:lyr:iassed.

The LCO ar:ir:il i es to the or:ierati ng l:lyr:iass removal feature only. If the l:lyr:ias~ enal:lle FtJriction is failed so as to r:irevent enteririg a l:lyr:iass condition, or:ieration may continue .*

The operatirig l:lypass settirigs are l:lased on analysis requirements for the l:lypassed functioris. These are discussed al:leve as part of the LCO discussiori for the affected Functioris.

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RPS Instrumentation B 3.3.l BASES APPLICABILITY This LCO requires all safety related trip functions to be OPERABLE is applicable in accordance with Table 3.3.1-1.

MODES 1 aRd 2, aRd iR MODES 3, 4, aRd 5 wheR more thaR oRe coRtrol rod is capable of beiRg withdrawR aRd PCS boroR COHCeRtratioH is less thaH REFUELING BORON CONCENTRATION.

While in MODES 3, 4, or 5, if PCS boron concentration is .at REFUELING BORON CONCENTRATION, or if no more than one full-length control rod is capable of being withdrawn, the RPS Function is already fulfilled (the safety analyses and the SHUTDOWN MARGIN definition both use the assumption that the highest worth withdrawn full-length control rod will fail to insert on a trip) and the safety analyses assumptions and SHUTDOWN MARGIN requirements will be met without the RPS trip Function.

As iHdicated iR Note (c) of Table 3.3.1 l, The High Startup Rate Trip Function is required to be OPERABLE in MODES 1 and 2, but may be bypassed when the .

associated wide range NI channel indicates below lE-4%

power, when poor counting statistics may lead to erroneous indication. It may also be bypassed when THERMAL POWER is above 13% RTP, where moderator temperature coefficient and fuel temperature coefficient make high rate of change of power unlikely. In MODES 3, 4, 5, and 6, the High Startup Rate trip is not required to be OPERABLE. Wide range channels are required to be OPERABLE in MODES 3, 4, and 5, by LCO 3.3.9, "Neutron Flux Monitoring Channels," and in MODE 6, by LCO 3.9.2, "Nuclear Instrumentation."

Tthe Loss of Load trip is ftO-t-required to be OPERABLE with THERMAL POWER at or above 17% RTP. Below 17% RTP, the ADVs are capable of relieving the pressure due to a Loss of Load event without challenging other overpressure protection.

The trips are designed to take the reactor subcritical, maintaining the SLs during AOOs and assisting the ESF in providing acceptable consequences during accidents.

If PCS boroH CORCeHtration is at REFUELING BORON CONCENTRATION, or if HO more thaH one coHtrol rod is capable of being withdrawH, the RPS Function is already fulfilled (the safety analyses aRd the SllUTDOWN P1ARGIN defi ni ti on both use the assumption that the highest worth withdrawR coRtrol rod will fail to iHsert on a trip) and the safety analys~s assumpti OHS aHd SllUTDOW~ MARGIN requi remeRtS wi 11 be met

without the RPS trip FunctioH .

Palisades Nuclear Plant B 3.3.1-27 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS The most commoH causes of chaHHel iHoperability are failure or drift of the bistable or process module sufficieHt to exceed the Al 1owabl e Value. Ty pi cal 1y, the drift is fouHd to be small aHd results iH a delay of actuatioH rather thaH a total loss of fuHctioH. This determiHatioH is geHerally made duriHg the performaHce of a CllANNEL FUNCTIONAL TEST wheH the process iHstrumeHt is set up for adjustmeHt to briHg it to withiH specificatioH. If the trip setpoiHt is less coHservative thaH the Allowable Value iH Table 3.3.1 l, the chaHftel is declared iHoperable immediately, aHd the appropriate CoHditioH(s) must be eHtered immediately.

The most common causes of channel inoperability are outright failure of -loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the "safe" condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically,

the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards).

In the event a channel *s trip setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or RPS bistable trip unit is found inoperable, -t-hetr-all affected Functions provided by that channel must be declared inoperable, and the plant must enter the Condition. for the particular protection Functions affected.

When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with' the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 is immediately entered if applicable in the current MODE of operation .

Palisades Nuclear Plant B 3.3.1-28 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS A Note has been added to the ACTIONS to clarify the .

(continued) application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function. The Completion Times of each inoperable Function will be tracked separately for each Function, starting from the time the Condition was entered.

Condition A applies to the failure of a single channel in any required RPS automatic trip Function, except High Startup Rate, ttttti Loss of Load, or ZPM Bypass Removal. RP5-coirwiderwe logic is rrnrmally two out of four. (Condition A is modified by a Note stating that this CoDdition does not apply to the High Startup Rate, ftf'ttt Loss of Load, or ZPM I Bypass Removal Functions. The failure of one channel of I those Functions is addressed by Conditions B, C, or D.) I If one RPS bistable trip unit or associated instrument * ,

channel is inoperable, startup or power operation is allowed to continue. Since the trip unit and associated instrument channel combine to perform the trip function, this Condition is also appropriate if both the trip unit and the associated instrument channel are inoperable. Though not explicitly required, the inoperable channel may should be ft-f"-i-p I*

channel) bypassed or tripped. If it is neither bypassed nor tripped, leaving the tnoperable trip Function in an untripped condition, administrative controls are provided to pre*1ent i naclvertent (trip cfiarrnel) bypass of the same Function in aHother chaHnel. The provision of four trip channels allows one channel to be* bypassed (removed from service) during operations, placing the RPS in

two-out-of-three coincidence logic. It is preferable to plac~ an inoperable chanHel in (trip chanRel) bypass rather than trip, siHce HO additional raHdom failure of a single channel can either spuriously trip the reactor or prevent it from tripping.

The failed channel must be +s-restored to OPERABLE status or

ts- placed in trip within 7 days. Restoring the channel to OPERABLE status restores the full capability of the Function. '

Required Action A.I places the Function in a one-out-of-three configuration. In this configuration,

common cause failure of dependent channels cannot prevent trip.

Palisades Nuclear Plant B 3.3.1-29 01/20/98 05/30/99

RPS Instrumentation

BASES ACTIONS A.1 (continued)

B 3.3.1 The Completion Time of 7 days is based on operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.

Condition B applies to the failure of a single High Startup Rate trip unit or associated instrument channel.

If one trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to entering MODE 2 from MODE 3. A shutdown provides the appropriate opportunity to repair the trip function and conduct the necessary testing. The Completion Time is based on the fact that the safety analyses take no credit for the functioning of this trip .

c.1 Condition C applies to the failure of a single Loss of Load or associated instrument channel.

If one trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to THERMAL POWER

~ 17% RTP following a shutdown. If the plant is shutdown at the time the channel becomes inoperable, then the failed channel must be restored to OPERABLE status prior to THERMAL POWER ~ 17% RTP. For this Completion Time, "following a shutdown" means this Required Action does not have to be completed until prior to THERMAL POWER ~ 17% RTP for the first time after the plant has been in MODE 3 following entry into the Condition. The Completion Time is based on the fact that the safety analyses take no credit for the functioning of this trip .

Palisades Nuclear Plant B 3.3.1-30 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS D.1 and D.2 (continued)

Condition D applies when one or more automatic ZPM Bypass removal channels are inoperable. If the ZPM Bypass removal channel cannot be restored to OPERABLE status, the affected ZPM Bypasses must be immediately removed, or the bypassed RPS trip Function channels must be immediately declared to be inoperable. Unless additional circuit failures exist, the ZPM Bypass may be removed by placing the associated 11 Zero Power Mode Bypass 11 key operated switch in the nonnal position.

A trip channel which is actually bypassed, other than as

.allowed by the Table 3.3.1-1 footnotes, cannot perfonn its specified safety function and must immediately be declared to be inoperable.

D.1 arHl D. 2 Condition D applies to one or two automatic (operating) bypass removal chan~els inoperable. If the bypass removal channel for any operating bypass cannot be restored to OPERABLE status, the associated RPS instrument channel may be considered OPERABLE only if the bypass is not in effect.

Otherwise, the affected RPS channel must be declared.

inoperable, and the bypass either removed or the bypass removal channel repaired. This is addressed by requirinl entry into the appropriate CoRdi ti on for the channels rendered inoperable by the bypass channel failure.

EB- .1 and EB-. 2 Condition EB- applies to the failure of two channels in any RPS a1:1tomatic trip Function, except lligh Startup Rate and Loss of Load ZPM Bypass Removal Function. (The failure of ZPM Bypass Removal Functions is addressed by Condition D.).

Co.ndition B-E is modified by a Note stating that this Condition does not apply to the lligh Startup Rate and Loss of Load FunctionsZPM Bypass Removal Function.

The Required Actions are ts-;nodified by a Note stating that LCO 3.0.4 is not applicable. The Note was added to allow*

the changing of MODES,even though two channels are inoperable, with one channel (trip channel) bypassed and one tripped. MODE changes in this configuration are allowed I because two trip channels for the affected function remain I OPERABLE. A trip occurring in either or both of those I channels would cause a reactor trip. I Palisades Nuclear Plant B 3.3.1-31 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS EG.1 and EG.2 (continued)

While it is conceptually possible that, if the two operable channels were those that do not have total channel separation in their cable routings, a single failure could disable both from tripping, in reality, such failures are extremely unlikely. Most failures involving a common cable fault would cause the affected channel (s) to fail in the de-energized condition, thereby initiating a reactor trip not preventing one. to permit maiRteRaRce aRd testiR~ oR oRe of the iRoperable chaRRels. In this configuration, the

protection system is in a one-out-of-two logic, and the probability of a common cause failure affecting both of the OPERABLE channels during the 7 days permitted is remote.

Required Action EG.1 provides for placing one inoperable channel in trip within the Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Though not explicitly required, the other inoperable channel may should be (trip channel) bypassed. If it is not bypassed, leaviRg oRe iRoperable trip FunctioR iR aR uRtripped coRditioR, admiRistrative coRtrols are provided to preveRt iRadverteRt bypass of the same FunctioR iR aRother chaRRel. *Such iRadverteRt bypass could defeat three of the four RPS chaRRels, reRderiRg the RPS inoperable.

This Completion Time is ~ufficient to allow the operator to take all appropriate actions for the failed channels while ensuring that the risk involved in operating with the failed channels is acceptable. With one channel of protective

. instrumentation bypassed or inoperable in an untripped condition, the RPS is in a two-out-of-three logic for that function; but with another channel failed, the RPS may be operating in a two-out-of-two logic. This is outside the assumptions made in the analyses and should be corrected.

To correct the problem, one of the inoperable channels is placed in trip. This places the RPS in a one-out-of-two for that function logic. If any of the other unbypassedOPERABLE channels for that function receives a trip signal, the

reactor will trip.

Action E.2 is modified by a Note stating that this' Action does not apply to (is not required for) the High Startup Rate and Loss of Load Functions .

Palisades Nuclear Plant B 3.3.1-32 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS EB.l and EB.2 (continued)

One channel is required to be restored to OPERABLE status within 7 days for reasons similar to those stated under Condition A. After one channel is restored to OPERABLE status, the provisions of Condition A still apply to the remaining inoperable channel. Therefore, the channel that is still inoperable after completion of Required Action EB.~

must be placed in trip if more than 7 days have elapsed since the initial channel failure.

The power range excore channels detectors are used to generate the internal ASI signal used as an input to the TM/LP trip. They ftftt! also provide input to the Thermal Margin Monitors for determination of the Q Power input to for the TM/LP trip and the VHPT. If two power range excore channels cannot be restored to OPERABLE status, power is restricted or reduced during subsequent operations because of increased uncertainty associated with inoperable power range excore channels which provide input to those trips.

The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is adequate to reduce power in an orderly manner without challenging plant systems.

Cor=iditior=i E applies to the failure of a sir=igle lligh Startup Rate trip ur=iit or associated ir=istrumer=it char=ir=iel. Sir=ice the trip ur=iit ar=id associated ir=istrumer=it char=ir=iel combir=ie to perform the trip fur=ictior=i, this Cor=iditior=i is also appropriate if both the trip ur=iit ar=id the associated ir=istrumer=it char=ir=iel are ir=ioperable. RPS coir=icider=ice,logic is r=iormally two out of four.

If one trip ur=iit or associated ir=istrumer=it char=ir=iel fails, it must be restored to OPERABLE status prior to er=iterir=ig MODE 3 from MODE 4. A shutdowr=i provides the appropriate opportur=iity to repair the trip fur=ictior=i ar=id cor=iduct the r=iecessary testing. The Completior=i*Time is based or=i the fact that the safety ar=ialyses take r=io credit for the fur=ictior=iir=ig

'of this trip. Ir=i additior=i, the probability of post mair=iter=iar=ice testir=ig of the restored char=ir=iel resultir=ig ir=i a

reactor trip is ur=iacceptable if performed at power sir=ice the Fur=ictior=i is r=iot credited.

Palisades Nuclear Plant B 3.3.1-33 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS .F.1 (continuedj CoRditioR F applies to the failure of a siRgle Loss of Load or associated iRstrumeRt chaRRel. SiRce the trip uRit aRd associated iRstrumeRt chaRRel combiR~ to perform the trip fuRctioR, this CoRditioR is also appropriate if both the trip uRit aRd the associated iRstrumeRt chaRRel are iRoperabl e. RPS coi Rei deRce logic is Rormall y two out of four.

If oRe trip uRit or associated iRstrumeRt chaRRel fails, it must be restored to OPERABLE status prior to Tl IERMAL POWER

~ 17% RTP followiRg a shutdowR. If the plaRt is shutdowR at the time the chaRRel becomes iRoperable, theR the failed chaRRel must be restored to OPERABLE status prior to TllERMAL POWER > 17% RTP. For this CompletiOR Time, "followiRg a shutdowR" meaRs this Required ActioR does Rot have to be completed URti l prior to TllERMAL POWER > 17°1!' RTP for the first time after the plaRt has beeR iR MODE 4 or 5 followiRg eRtry iRto the CoRditioR. The CompletioR Time is based OR the fact that the safety aRalyses take RO credit for the fuRctioRiRg of this trip. IR additioR, the probability of post maiRteRaRce testiRg of the restored chaRRel resultiRg iR a reactor trip is uRacceptable if performed at power siRce'the FuRctioR is Rot credited.

6.1 aRd 6.2 CoRditi OR 6 applies to two Iii gh Startup Rate trip uRits or associated iRstrumeRt chaRRels iRoperable. SiRce the trip uRit aRd associated iRstrumeRt chaRRel combiRe to perform the trip fuRctioR, this CoRditioR is also appropriate if two trip URits aRd their associated iRstrumeRt chaRRels are iRoperable. Required ActioR 6.1 requires placiRg oRe iRoperable chaRRel iR trip withiR 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Though Rot explicitly required, the other iRoperable chaRRel should be (trip chaRRel) bypassed. If it is Rot bypassed, oRe iRoperable trip FURCtioR is iR aR URtripped CORditiOR aRd admiRistrative coRtrols are provided to preveRt iRadverteRt bypass the same FuRctioR iR aRother chaRRel. Such aR iRadverteRt bypass could defeat three of the four RPS chaRRels, thereby reRderiRg the RPS iRoperable .

Palisades Nuclear Plant B 3.3.1-34 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS G.1 and G.2 (continued)

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is sufficient to allow the operator to take all appropriate actions for the failed channels whil~ minimizing the risk involved in operating with the failed channels. With one channel bypassed or inoperable in an untripped condition, the RPS is in a two out of three logic, but with another channel failed, the RPS may be operating in a two out of two logic. One of the channels must be placed in trip to place the RPS in a one out of two logic. If any of the other OPERABLE channels receives a trip signal, the reactor will trip.

Required Action G.2 requires one channel to be restored to OPERABLE status prior to entering MODE 3 from MODE 4 for reasons similar to those stated under Condition E. After one channel is restored to OPERABLE status, the provisions of Condition E still apply to Hie remaining inoperable channel. Therefore, the channel that is still inoperable must also be restored to OPERABLE status prior to entering MODE 3 from MODE 4.

11.1 and 11.2 Condition II applies to two Loss of Load trip units or associated instrument channels inoperable. Since the trip unit and associated instrument channel combine to perform the trip function, this Condition is also appropriate if two trip units and their associated instrument channels are inoperable. Required Action 11.1 requires placing one inoperable channel in trip within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Though not explicitly required, the other inoperable channel should be (trip channel) bypassed. If it is not bypassed, one inoperable trip Function is in an untripped condition and administrative controls are provided to prevent inadvertent bypass of the same Function in another channel. Such an

inadvertent bypass could defeat three of the four RPS channels, thereby rendering the RPS inoperable.

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is sufficient to allow the operator to take all appropriate actions for the failed channels while minimizing the risk involved in operating with the failed channels. With one channel bypassed or inoperable in an untripped condition, the RPS is in a two out of three logic, but with another channel failed, the RPS may be operating in a two out of two logic. One of the channels must be placed in trip to place the RPS in a one out of two logic. If any of the other OPERABLE channels receives a trip signal, the reactor will trip.

Palisades Nuclear Plant B 3.3.1-35 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES ACTIONS G.1 and G.2 (continued)

Ref!ui red Ac ti on II. 2 ref!ui res one channel to be restored to OPERABLE status prior to TllERMAL POWER ~ 17°~ RTP following a shutdown for reasons similar to those stated under Condition F. After one channel is restored to OPERABLE status, the provisions of Condition F still apply to the remaining inoperable channel. Therefore, the channel that is still inoperable must also be restored to OPERABLE status prior to TllERMAL POWER 2 17°~ RTP foll o*ning a shutdown.

G.1. G.2.1. and G.2.2I.1. 1.2.1 and 1.2.2 I Condition Gt is erytered when the Required Action and associated Completion Time of Condition A, B, C, D, E, or F~

G or II are not met, or if the control room ambient air temperature exceeds 90°F.

If the control room ambient air temperature exceeds 90°F*,

all Thermal Margin Monitor channe.ls are rendered inoperable

\

because their environmental qualification temperature limit is exceeded. In this condition, or if the Required Actions and associated Completion Times are not met, the reactor

. must be placed in a condition in whi*ch the LCO does not apply. To ~ccomplish this, the plant must be placed in MODE 3, with no more than one full-length control rod capable of being withdrawn or with the PCS boron concentration at REFUELING BORON CONCENTRATION in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

. The Completion Time is reasonable, based on operating experience, for placing the plant in MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The Completion Ti.me is also reasonable to ensure that no more than one full-length control rod is capable of being withdrawn or*that the PCS boron concentration is at REFUELING BORON CONCENTRATION.

SURVEILLANCE The SRs for any particular RPS Function are found in*the SR REQUIREMENTS column of Table 3.3.1-1 for that Function. Most Functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONA~ TEST, and CHANNEL CALIBRATION.

While Palisades is not committed to performing all testing I discussed in ANSI/IEEE Standard 338-1977, CHANNEL CHECKS, I

CHANNEL FUNCTIONAL TESTS, AND CHANNEL CALIBRATIONS are performed in accordance with the guidance of ANSI/IEEE Standard 338-1977, which is endorsed by Regulatory Guide 1.118.

I I

, I I

Palisades Nuclear Plant B 3.3.1-36 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. Under most conditions, a CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including inditation and readability. If a channel is

. outside the criteria, it may be an indication that the

transmitter or the signal processing equipment has drifted

outside its limits.

The Containment High Pressure and Loss of Load channels are pressure switch actuated~ As such, they have no associ~ted control room indicator and do not require a CHANNEL CHECK.

The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure.

Since the probability of two random failures in redundant channels in any 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels.

SR 3.3.1.2 This SR verifies that the control room ambient air temperature is within the environmental qualification temperature limits for the most restrictive RPS components, which are the Thermal Margin Monitors. These monitors provide input to both the VHPT Function and the TM/LP Trip Function. The 12 hout Frequency is reasonable based on engineering judgement and operating experience .

Palisades Nuclear Plant B 3.3.1-37 oi/20/98 os/30/99

RPS Instrumentation B 3.3.l B.ASES SURVEILLANCE SR 3.3.1.3 REQUIREMENTS (continued) A daily calibration (heat balance) is perfonned when THERMAL POWER is ~ 15%. The daily calibration consists of adjusting the "nuclear power calibrate" potentiometers to agree with the calorimetric calculation if the absolute difference is ~ 1.5% ~- Nuclear power is adjusted via a potentiometer, or THERMAL POWER is adjusted via a Thennal Margin Monitor bias number, as necessary, in accordance with the daily calibration (heat balance) procedure. Perfonnance of the daily calibration ensures that the two inputs to the Q power measurement are indicating accurately with respect to the much more accurate secondary calorimetric calculation.

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on plant operating experience and takes into account indications and alanns located in the control room to detect deviations in channel outputs.

The Frequency is modified by a Note indicating this

Surveillance must be perfonned within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is ~ 15% RTP. The secondary calorimetric is inaccurate at lower power levels. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .allows time requirements for plant stabilization, data taking, and instrument calibration.

SR 3.3.1.4 It is necessary to cempare aHd, if Hecessary, calibrate the power range excore channel upper and lower subchannel amplifiers such that the iHterHal measured ASI used iH the TM/LP trip reflects the true core power distribution as detennined by the incore detectors. ASI is utilized as an input to the TM/LP trip function where it is used to ensure that the measured axial power profiles are bounded by the axial power profiles* used in the development of the T;niet limitation of LCD 3.4.1. An adjustment of the excore channel is necessary only if individual excore channel measured ASI is cempared te the tetal cere differs from AXIAL OFFSET, as measured by the incores, aHd the differeHce f-5" by greater than 0.02.

A Note te the Fre~ueHcy indicates the Surveillance is not required to have been perfonned until withiH 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is ~ 25% RTP. *Uncertainties in the excore and incore measurement process make it impractical to calibrate when THERMAL POWER is < 25% RTP. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allows time for plant stabilization, data taking, and instrument calibration.

Palisades Nuclear Plant B 3.3.1-38 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.4 (continued)

REQUIREMENTS The 31 day Frequency is adequate, based on operating experience of the excore linear amplifiers and the slow burnup of the detectors. The excore readings are a strong function of the power produced in the peripheral fuel bundles and do not represent an*integrated reading across the core. Slow changes in neutron flux during the fuel cycle can also be detected at this Frequency.

SR 3.3.1.5 A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel. except Loss of Load and High Startup Rate. every 92 days to ensure the entire channel will perform its intended function when needed. For the TM/LP Function. the constants associated with the Thermal Margin Monitors must be verified to be within tolerances.

A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Bistable Tests The bistable setpeint must be feune te trip within the Allewable Values specifiee iR the LCO aRe left set censistent with th~ assumptieRs ef the setpeint aRalysis.

A test ~i~nal is superimpesee eR the iRput in ene chanRel at a time te verify that the bistable trips HithiR the specifiee telerance areuRe the setpeiRt. This is eeRe with the affeetee RPS chaRRel (trip chaRRel) bypassee. Any setpoint adjustment must be consistent with the assumptions of the current setpoint analysis.

The as feuRe aRe as left values must also be recoreee aRe reviewee for coRsisteRcy with the assumptieRs of the FrequeRcy exteRsieR analysis. The requiremeRts for this review are outliRee iR RefereRce 5 .

The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327. "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5).

Palisades Nuclear Plant B 3.3.1-39 01/20/98 05/30/99

- BASES RPS Instrumentation B 3.3.1 SURVEILLANCE SR 3.3.1.6 REQUIREMENTS (continued) A calibration check of the power range excore channels using the internal test circuitry is required every 92 days. This SR uses an internally generated test signal to check that the 0% and 50% levels read within limits for both the upper and lower detector, both on the analog meter and on the TMM screen. This check verifies that neither the zero point nor the amplifier gain adjustment have undergone excessive drift since the previous complete CHANNEL CALIBRATION.

The as fouRd aRd as left values must also be recorded aRd reviewed for coRsisteRcy with the assumptioRs of the frequeRcy exteRsioR aRalysis. The requiremeRts for this review are outliRed iR Refereftce 5.

The R~utroft detectori are excluded from dalfbratioR because they are passive devices with miRimal drift aRd because of the difficulty of simulatiRg a meaRiRgful sigRal. Slow chaRges iR detector seRsitivity are compeRsated for by

performiRg the daily calorimetric calibratioR (SR 3.3.1.3) aRd the moRthly .liRear subchaHRel gaiR check (SR 3.3.1.4).

IR additioR, associated coRtrol_room iRdicatioRs are coRtiRuously moRitored by the operators.

The Fre~uency of 92 days is ~cceptable, based on.plant operating experience, and takes into account indications and alarms available to the operator in the control room.

SR 3.3.1.7 A CHANNEL FUNCTIONAL TEST on the Loss of Load and High Startup Rate channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required.

A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Palisades Nuclear Plant B 3.3.1-40 01/20/98 05/30/99

l RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.7 (continued)

REQUIREMENTS The High Startup Rate trip is actuated by either of the Wide Range Nuclear Instrument Startup Rate channels. NI-61/63 sends a trip signal to RPS channels A and C; NI-62/64 to channels B and D. Since each High Startup Rate channel would cause a trip on two RPS channels, the High Startup

Rate trip is not tested when the reactor is critical .--ffte lligh Startup Rate trip Function is requiree euring startup operation ane may be operationally bypassee when below lE 4~

RTP or above 13% RTP.

The four Loss of Load Trip channels are all actuated by a single pressure switch monitoring turbine auto stop oil pressure which is not tested when the reactor is critical.

Operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once per 7 days prior to each reactor startup.

SR 3.3.1.8 SR 3.3.1.8 is the performance of a CHANNEL CALIBRATiON every 18 months.

CHANNEL CALIBRATION is a complete check of the instrument channel .including the sensor (except neutron detectors).

The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performee consistent with the plant specific setpoint analysis.

The bistable setpoints must be found to trip within the Allowable Values specified in the LCO and left set consistent with the assumptions of the setpoint analysis.

The Variable High Power Trip setpoint shall be verified to reset properly at several indicated power levels during (simulated) power increases and power decreases.

The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the setpoint analysis. frequency extension analysis. The requirements for this review are outlinee in Reference 5 .

Palisades Nuclear Plant B 3.3.1-41 01/20/98 05/30/99

RPS Instrumentation

BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.8 (continued)

B 3.3.1 As part of the CHANNEL CALIBRATION of the wide range Nuclear Instrumentation, automatic removal of the rPMG ZPM Bypass for the Low PCS Flow, TM/LP, and Lew SG Pressure trips and the automatic (eperatienal) bypassing ef the.Less ef Lead and Iii gh Startup Rate trips must be. verified to assure that these trips are available when required.

The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift.

This SR is modified by a Note which states that it is not I necessary to calibrate neutron detectors because they are *I passive devices with minimal drift and because of the I difficulty of simulating a meaningful signal. Slow changes I in power range excore neutron detector sensitivity are I compensated for by performing the daily calorimetric I calibration (SR 3.3.1.3) and the monthly calibration using I the incore detectors (SR 3.3.1.4). Sudden changes in I

detector per\ormance would be noted during the required CHANNEL CHECKs (SR 3.3.1.1). .

The neutron detectors are excluded frem CllANNEL CALIBRATION because they are passh*e de*vi ces with minimal drift and because ef the difficulty ef simulating a meaningful signal.

I I

Slew changes in detector sensitivity are compensated fer by performing the daily calorimetric calibration (SR 3.3.L3) and the monthly linear subchannel gain check (SR 3.3.1.4).

REFERENCES 1. 10 CFR 50, Appendix A, GDC 21

2. 10 CFR 100

3. IEEE Standard 279-1971, April 5, 1972

4. FSAR, Chapter 14

5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989 Palisades Nuclear Plant B 3.3.1-42 01/20/98 05/30/99

RPS Instrumentation B 3.3.1 Table B 3.3.1-1 (page 1 of 1)

Instruments.Affecting Multiple Specifications REQUIRED AFFECTED INSTRUMENT CHANNELS SPECIFICATIONS Source Range NI-1/3 & 2/4, Count Rate Signal 3.3.9 3.9.2 Source Range NI-1/3, Count Rate Indication @ C-150 Panel 3.3.8 #1 Wide Range Nl-1/3 & 2/4, Flux Level 10~ Bypass 3.3.1 #3,6,7,9,&12 Wide Range NI-1/3 & 2/4, Startup Rate 3.3.1 #2 Wide Range NI-1/3 & 2/4, Flux Level Indication 3.3.7 #3 3.3.9 Power Range NI-5, 6, 7 & 8, Tq 3.2.1 3.2.3 Power Range Nl-5~ 6, 7 & 8, Q Power 3.3.1 #1 & 9 Power Range NI-5, 6, 7 & 8, ASI 3.3.1 #9 3.2.1 3.2.4 Power Range NI-5, 6, 7 & 8, Loss of Load/High Startup Rate Bypass 3.3.1 /12 & 10 PCS TC TT-0112 & 0122 CC & CD, Temperature Signal (SMM) 3.3.7 #5 PCS TC TT-0112 & 0122 CA, CB, CC & CD, Temperature Signal (Q Power & TMM) 3.3.1 #1 & 9 3.4.1.b PCS TC TT-0112CA & 0122CB, Temperature Signal (LTOP) 3 .4.12 .b..l PCS TC TT-0112CC & 0122CD (PTR-0112 & 0122) Temperature Indication 3.3.7 #2 PCS TC TT-0112CA, Temperature Signal (SPI 6T Power for PDIL Alarm Circuit) 3.1.6 PCS TC TT-0122CB, Temperature Signal (PIP 6T Power for PDIL Alarm Circuit) 3.1.6 PCS TH TT-0112 & 0122 HC & HD, Temperature Signal (SMM) 3.3.7 #5 PCS TH TT-0112HC & 0122HD (PTR-0112 & 0122) Temperature Indication 3.3.7 #1 PCS TH TT-0112 & 0122 HAi HB, HC & HD, Temperature Signal (Q Power) 3.3.1 #1 & 9 PCS TH TT-0112HA, Temperature Signal (SPI 6T Power for PDIL Alarm Circuit) 3.1.6 PCS TH TT-0122HB, Temperature Signal (PIP 6T Power for PDIL Alarm Circuit) 3.1.6 Thermal Margin Monitor PY-0102A, B, C, & D 3.3.1 #1 & 9 Pressurizer Pressure PT-0105A & B, Pr~ssure Signal (WR Indication & LTOP) 3.3.7 #5 3.3.4.12.b.l Pressurizer Pressure PT-0102A, B, C & D, Pressure Signal (RPS & SIS) 3.3.1 #8 & 9 3.3.3 #1.a Pressurizer Pressure PT-0104A & B, Pressure Signal (LTOP & SOC Interlock) 3.4.12.b.1 3.4.14 Pressurizer Pressure PI-0110, Pressure Indication @ C-150 Panel 3.3.8 #2 SG Level LT-0751 & 0752 A, B, C & D, Level Signal (RPS & AFAS) 3.3.1 #4 & 5 3.3.3 #6.a & 6.b SG Level LI-0757C & 0758C, Wide Range Level Indication @ C-150 Panel 3.3.8 #10 & 11 SG Level LI-0757 & 0758 A & B, Wide Range Level Indication 3.3.7 #11 & 12 SG Pressure PT-0751 & 0752 A, B, C&D, Pressure Signal (RPS & SG Isolation) 3.3.1 #6 & 7 3.3.3 #4.a & 4.b SG Pressure PIC-0751 & 0752 A, B, C & D, Pressure Indication 3.3.7 #13 & 14 SG Pressure PI-0751E & 0752E, Pressure Indication @ C-150 Panel 3.3.8 #8 & 9 Containment Pressure PS-1801, 1802, 1803&1804, Switch Output (RPS) 3.3.1 #11 Containment Pressure PS-1801, 1802A, 1803 & 1804A, Switch Output (ESF Actuation) 3.3.3 112.a Containment Pressure PS-l801A, 1802, 1803A & 1804, Switch Output (ESF Actuation) 3.3.3 #2.b

Note: The information provided in this table is intended for use as an aid to distinguish those instrument channels which provide more than one required function and to describe which specifications they affect. The information in this table should not be taken as inclusive for all instruments nor affected specifications.

Palisades Nuclear Plant B 3 .3 .1-43 01/20/98 05/30/99

RPS Logic and Trip Initiation

B 3.3 INSTRUMENTATION B 3.3.2 Reactor Protective System (RPS) Logic and Trip Initiation B 3.3.2 BASES BACKGROUND The RPS initiates a reactor trip to protect against violating the eere speeifiee acceptable fuel design limits and reactor coolant pressure boundary integrity during Anticipated Operational Occurrences (AOOs). *By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.

The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters anq equipment performance.

The LSSS, defined in this Specification as the Allowable Value, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits

during Design Basis Accidents (DBAs) .

During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:

The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling;

Fuel centerl*ine melting shall not occur; and

The Primary Coolant System (PCS) pressure SL of 2750 psia shall not be exceeded.

Maintaining the parameters within the above values ensures that the offsite dose will be within the 10 CFR 50 (Ref. 1) and 10 CFR 100 (Ref. 2) criteria during AOOs.

Accidents are events that are analyzed even though they are not expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall b~ maintained within an acceptable fraction of 10 CFR 100 (Ref. 2) limits.

Different accident categories allow a different fraction of these limits based on,probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event .

Palisade~ Nuclear Plant B 3.3.2-1 01/20/98 05/30/99

RPS Logic and Trip Initiation

BASES BACKGROUND The RPS is segmented into four interconnected modules.

B 3.3.2 (continued) These modules are:

Measurement channels (or pressure switches);

Bistable trip units;

Matrix Logic; and

Trip Initiation Logic.

This LCD addresses the RPS Logic (Matrix Logic and Trip Initiation Logic), including Manual Trip capability.

LCD 3.3.1, "Reactor Protective System (RPS) Instrumentation,"

provides a description of the role of the measurement channels and associated bistable trip units in the RPS. The RPS Logic.

is summarized below:

RPS Logic

The RPS Logic, consisting of Matrix Logic and Trip Initiation Logic, employs a scheme that provides a reactor trip when trip units in any two of the four channels sense the same input parameter trip. This is called a two-out-of-four trip logic.

This logic and the clutch power supply configuration are shown

. in FSAR Figure 7-1 (Ref. 3). .

Bistable trip unit relay contact outputs from the four channels are configured into six logic matrices. Each logic matrix checks for a coincident trip in the same parameter in two trip unit channels. The matrices are designated the AB, AC, AD, BC, BD, and CD matrices to reflect the bistable trip unit channels being monitored. Each logic matrix contains four normally energized matrix relays. When a coincidence is detected, consisting of a trip in the same Function in the two channels being monitored by the logic matrix, all four matrix relay coils de-energize.

The matrix relay contacts are arranged into trip paths, with one of the four matrix relays in each matrix opening contacts in one of the four trip paths. Each trip path provides power to one o,f the four nonnally energized clutch power supply 11 M-contactors11 (Ml, M2, M3, and M4). The trip paths thus each have six contacts in series, one from each matrix, and perform a logical OR function, de-energizing the M-contactors if any one or more of the six logic matrices indicate a coincidence condition.

Palisades Nuclear Plant B 3.3.2-2 01/20/98 05/30/99

RPS Logic and Trip Initiation B 3.3.2 BASES BACKGROUND When a coincidence occurs in two RPS channels, all four

. (continued) matrix relays in the affected matrix de-energize. This in turn de-energizes all four M-contactors, which interrupt AC input power to the four clutch power supplies, allowing the*full-length control rods to insert by gravity.

Manual* reactor trip capability is afforded by two main control panelheitrti-mounted pushbuttons. One of these (on Control Panel C0-2) opens contacts in series with each of the four trip paths, de-energizing all M-contactors .. The other pushbutton (on Control Panel C0-6) opens circuit breakers which provide AC input power to the Mcontactor contacts and downstream clutch power supplies. Thus depressing either pushbutton will cause a reactor trip.

De-energizing the M-contactors removes AC power to the four clutch power supply inputs. Contacts from M-contactors Ml and M2 are in series with each other and in the AC power supply path to clutch power supplies PSI and PS2 (these constitute a "trip leg"). M3 and M4 are similarly arranged with respect to clutch power supplies PS3 and PS4 (these constitute a second "trip leg"). Approximately half of the control rod~ clutches receive power from auctioneered clutch power supplies 1 and 3.

The remaining control rod~ clutches receive clutch power from auctioneered clutch power supplies 2 and 4.

Matrix .Logic refers to the matrix power supplies, trip channel bypass contacts, and intercpnnecting RPS cabnet .matrix wiring between bistable and auxiliary trip unit relay contacts, ttp-te Btlt net including the matrix relays. Contacts in the bistable and auxiliary trip units are excluded from the Matrix Logic definition, since they are addressed as part of the meastlrement instrumentation channel.

The Trip Initiation Logic consists of the matrix relays and their asseciated centacts, the M-contactor isolation transfonners, all interconnecting wiring, CO 2 mantlal trip centacts, and the M-contactors.. The CO 6 mantlal trip is net part ef the Trip Initi~tien Legic, Btlt directly catlses epening ef the RPS ci rctli t Breakers which remev*es all AC pmver frem the cl tltch pewer s*tlppl i es.

Manual trip circuitry includes both manual reactor trip pushbuttons C0-2 and C0-6, and the interconnecting wiring necessary to effect de-energization of the clutch power

supplies .

Palisades Nuclear Plant B 3.3.2-3 01/20/98 05/30/99

l RPS Logic and Trip Initiation B 3.3.2 BASES BACKGROUND Neither the clutch power supplies nor the AC input power (continued) source to these supplies is considered as safety related other than as addressed by this LCD. Operation may continue with one or two selective clutch power supplies de-energized.

It is possible to change the two-out-of-four RPS Logic to a two-out-of-three logic for a given input parameter in one channel at a_ time by ftTrip Cehannelt--bBypassing the RPS Trip I*

unit output contacts in the Matrix Logic "Ladder." select I portions of the Matrix Logic. Trip Channel Bhypassing a trip I unit effectively shorts the bistable trip unit relay contacts in the three matrices associated with that channel. Thus, the bypassed bistable trip units will function normally, producing normal channel trip indication and annunciation, but a reactor trip will not occur unless two additional channels indicate a

. trip condition. Trip Cehannel Bhypassing can be simultaneously performed on any number of parameters in any *number* of channels, providing each parameter is bypassed in only one channel at a time. A single bypass key for each trip function interlock prevents simultaneous ftTrip Cehannelt Bhypassing of the same parameter in more than one channel. Trip Cehannel Bhypassing is normally employed during maintenance or testing.

Functional testing of the entire RPS, from bistable trip unit input through the de-energizing opening of individual sets of clutch power supplies, can be performed either at power or during shutdown and is normally performed on a quarterly basis.

FSAR-. Sect1on 7.2 (Ref. 4) 1 explains RPS testing in more detail.

APPLICABLE Reactor Protective System (RPS) Logic SAFETY ANALYSES The RPS Logic provides for automatic trip initiation to avoid exceeding maintaiR the SLs during AOOs and to assist the -ESF systems in ensuring acceptable consequences during accidents.

All transients and accidents that call for a reactor trip assume the RPS Logic is functioning ~s designed .

Palisades Nuclear Plant 8 3.3.2-4 01/20/98 05/30/99

RPS Logic and Trip Initiation

- BASES Manual Trip B 3.3.2 APPLICABLE SAFETY ANALYSES (continued) There are no accident analyses that take credit for the Manual Trip; however, the Manual Trip is part of the RPS circuitry.

It is used by the operator to shut down the reactor whenever any parameter is rapidly trending toward its trip setpoint. A Manual Trip accomplishes the same results as any one of the automatic trip Functions.

The RPS Logic and Trip Initiation satisfy Criterion 3 of 10 CFR 50.36(c)(2).

LCO Reactor Protective System (RPS) Logic

~ai.lures of individual trip unit bistable relays and their contacts are addressed in LCO 3.3.1. This Specification addresses failures of the Matrix Logic not addressed in the above, such as the failure of matrix relay power supplies or the failure of the trip channel bypass contact in the bypass condition. **

Loss of a single vital bus will de-energize one of the two power supplies in each of three matrices. Because of power supply auctioneering, all four matrix relays will remain energized in each affected matrix. This de-energization of up to three matrix power supplies due to a single failure is to be treated as a single channel failure.

Each of the four Trip Initiation Logic channels de-energizes one set of clutch power supplies if any of the six coincidence.

matrices de-energize their associated matrix relays. They thus perfonn a logical OR function. Trip Initiation Logic channels 1 and 2 receive AC power from preferred AC bus Y-30. Trip Initiation Logic channels 3 and 4 receive AC input power fr*om preferred AC bus Y-40. Because of clutch power supply output auctioneering, it is possible to de-energize either input bus without de-energizing control rod clutches.

1. Matrix Logic This LCO requires six channels of Matrix Logic to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod is capable of being

withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION.

Palisades Nuclear Plant B 3.3.2-5 01/20/98 05/30/99

RPS Logic and Trip Initiation B 3.3.2 BASES LCO 2. Trip Initiation Logic (continued)

This LCO requires four channels of Trip Initiation Logic to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and ~

when more than one full-length control rod is capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION.

3. Manual Trip The LCO requires both Manual Trip channels to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than

one full-length control rod is capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION.

Two independent pushbuttons are provided. Each pushbutton is considered a channel. Depressing either pushbutton interrupts power to al.l four clutch power supplies, tripping the reactor.

APPLICABILITY The RPS Matrix Logic, Trip Initiation Logic, and Manual Trip are required to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod capable of I being withdrawn and the PCS boron concentration is less than ..

REFUELING BORON CONCENTRATION. This ensures the reactor can b~

tripped when necessary, but allows for maintenance and testing when the reactor trip is not needed.

In MODES 3, 4, and 5 With no more than one full-length control rod capable of being withdrawn or the PCS boron concentration at REFUELING BORON CONCENTRATION, these Functions do not have to be OPERABLE. However, LCO 3.3.9, "Neutron Flux Monitoring Channels," does require neutron flux monitoring capability under these conditions.

ACTIONS When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with the same trip Function; then the plant is outside the safety analysis. Therefore, LCO 3.0.3 is immediately entered if

applicable in the current MODE of operation .

Palisades Nuclear Plant B 3.3.2-6 01/20/98 05/30/99

RPS Logic and Trip Initiation B 3.3.2 BASES ACTIONS A.:...l (continued)

Condition A applies if one Matrix Logic channel is inoperable in any applicable MODE. Loss of a single preferred AC bus will de energize ene of the two matrix power supplies in up to three matrices. This is considered a single matrix failure.

The channel must be restored to OPERABLE status within

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provides the operator time to take appropriate actions and still ensures that any risk involved in operating with a failed channel is acceptable. Operating experience has demonstrated that the

probability of a random failure of a second Matrix Logic channel is low during any given 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months interval. If the channel cannot be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , Condition E is* entered .

Condition B applies if one Trip Initiation Logic channel is inoperable in any applicable MODE. The Required Action. require de-energizing the affected clutch power supplies. This removes the need for the affected channel by performing its associated safety function. With the clutch power supplies associated with one initiation logic channel de-energized, the remaining two clutch power supplies prevent control rod clutches from de-energizing. The remaining clutch power supplies are in a one-out-of-two logic with respect to the remaining initiation logic channels in the clutch power supply path. This meets redundancy requirements, but testing on the OPERABLE channels cannot be performed without causing a reactor trip.

Required Action 8.1 provides for de-energizing the affected clutch power supplies associated with the inoperable channel within a Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This Required Action is conservative, since the redundant initiation logic channel associated with the same set of clutch power will de energize the affected clutch power supplies if required during the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time.

Palisades Nuclear Plant B 3.3.2-7 01/20/98 05/30/99

RPS Logic and Trip Initiation B 3.3.2 BASES ACTIONS C.1 (continued)

Condition Capp.lies to the failure of one Manual Trip channel in any applicable MODE. With one manual reactor trip channel inoperable operation may continue until the reactor is shut down for other reasons. Repair during operatfon is not

required because one OPERABLE channel is all that is required for safe operation. No safety analyses assume operation of the Manual trip.

The Manual Trip channels are not testable without actually causing a reactor trip, so even if the difficulty were.

corrected, the post maintenance testing necessary to declare the channel OPERABLE could not be completed during operation.

Because of this, the Required Action is to restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3 during the next plant startup .

Condition D applies to the failure of both Trip Initiation Logic channels affecting the same trip leg. The affected control rod drive clutch power supplies must be de-energized*

immediately. With both channels inoperable, the RPS Function is lost if the affected clutch power supplies are not de-energized. Therefore, immediate.action is required to de-energize the affected clutch power supplies. The immediate Completion Time is appropriate since there could be a loss of safety function if the associated clutch power supplies are not de-energized. Entry into~LCO 3.0.3 is not an acceptable alternative in this condition.

E.1. E.2.1 and E.2.2 Condition E is entered if Required Actions associated with Condition A, B, C, or D are not met within the required Completion Time or if for one or more Functions more than one Manual Trip, Matrix Logic, or Trip Initiation Logic channel is inoperable for reasons other than Condition D.

'In Condition E the reactor must be placed in a MODE in which the LCO does not apply. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in MODE 3 is reasonable, based on operating experience, to

reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.

Palisades Nuclear Plant B 3.3.2-8 01/20/98 05/30/99

RPS Logic and Trip Initiation

BASES B 3.3.2 ACTIONS E.1. E.2.i and E.2.2 (continued)

Required Actions E.2.1 and E.2.2 allow 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to verify that no more than one full-length control rod is capable of being withdrawn or to verify that PCS boron concentration is at REFUELING BORON CONCENTRATION. The Completion Time is reasonable to place the plant in an operating condition in which the LCO does .not apply.

SURVEILLANCE SR 3.3.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST on each RPS Logic channel is performed every 92 days to ensure the entire channel will perform its intended function when needed. A successful test I of the required contact(s) of a channel relay may be performed I by the verification of the change of state of a single contact I of the relay. This clarifies what is an acceptable CHANNEL I FUNCTIONAL TEST of a relay. This is acceptable because all of I

the other required contacts of the relay are verified by othe~

Technical Specificati-0ns and non-Technical Specifications tests at least once per refueling interval with applicable ex tens i ans.

This SR addresses the two tests associated with the RPS Logic:

J I

I I

I Matrix Logic and Trip Initiation Logic.

Matrix Logic Tests These tests are performed one matrix at a time. They verify that a coincidence in the two input channels for each Function removes power from the matrix relays. During testing, power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. The Matrix Logic tests will detect any short circuits around the bistable contacts in the coincidence logic such as may be caused by faulty bistable relay or trip channel bypass contacts .

Palisades Nuclear Plant B 3.3.2-9 01/20/98 05/30/99

RPS Logic and Trip Initiation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.1 (continued)

REQUIREMENTS Trip Initiation Logic Tests These tests are similar to the Matrix Logic tests, except that test power is withheld from one matrix relay at a time, allowing the initiation circuit to de-energize, de-energizing the affected set of clutch power supplies.

The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). .

SR 3.3.2;2 A CHANNEL FUNCTIONAL TEST on the Manual Trip channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required.

A successful test of the required contact(s) of a channel relay I may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-I I

I I

I Technical Specifications tests at least once per refueling I interval with applicable extensions. *I The Manual Trip Function is not tested at power. However, the simplicity of this circuitry and the absence of drift concern makes this Frequency adequate. Additionally, operating experience has shown that these components usually pass the Surveillance when performed once within 7 days prior to each reactor startup.

REFERENCES 1. 10 CFR 50, Appendix A

2. 10 CFR 100

3. FSAR, Figure 7-1

4. FSAR, Section 7.2

5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989 Palisades Nuclear Plant B 3.3.2-10 01/20/98 05/30/99

ESF Instrumentation B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Engineered Safety Features (ESF) Instrumentation BASES BACKGROUND The ESF Instrumentation (ESFI) initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System {PCS) pressure boundary and to mitigate accidents.

The ESFI eefltaifls eeviees aflfl circuitry tfl-ttt--generates the signals listed below when the monitored variables reach levels that are indicative of conditions requiring protective action.

Alse listee are the The inputs to each ESF aActuation s£ignal are also listed.

1. Safety Injection Signal {SIS).

a. Containment High Pressure (CHP)

5r.

b. Pressurizer Low Pressure Containment High Pressure Signal {CHP);

a. Con~ainment High Pressure - Left Train

b. Containment High Pressure - Right Train 63-. Containment High Radiation Signal (CHR);

a. Containment High Radiation

24. Steam Generator Low Pressure {SGLP);

a. Steam Generator fStr)-A Low Pressure

b. Steam Generator B Low Pressure

35. Recirculation Actuation Signal (RAS);

a. Safety Injection Refueling Water Tank (SIRWT) Low Level

46. Auxiliary Feedwater Actuation Signal (AFAS);

a.

b.

Steam Generator A Low Level Steam Generator B Low Level Palisades Nuclear Plant B 3.3.3-1 01/20/98 05/30/99

ESF Instrumentation B 3.3.3 BASES BACKGROUND 7. Automatic Bypass Removal (continued)

a. Pressurizer Pressure Low Bypass

b. Steam Generator A Low Pressure Bypass

c. Steam Generator B Low Pressure Bypass In the above list of actuation signals, the CHP and RAS are derived from pressure and level switches, respectively.

Equipment actuated by each of the above signals is identified in the FSAR, Chapter 7. (Ref. 1).

The ESF circuitry, with the exception of RAS, employs two-out-of-four logic. Four independent measurement channels are provided for each function used to generate ESF actuation signals. When any two channels of the same function reach

their setpoint, actuating relays are energized which, in turn, initiate the protective actions. Two separate and redundant trains of actuating relays, each powered from separate power supplies, are utilized. These separate relay trains operate redundant trains of ESF equipment.

RAS logic consists of output contacts of the relays actuated by the SIRWT level switches arranged in a one-out-of-two taken 11 twice" logic. The contacts are arranged so that at least one low level signal powered from each station battery is required to initiate RAS. Loss of a single battery, therefore, cannot either cause or prevent RAS initiation.

The ESF logic circuitry contains the capability to manually block the SIS actuation logic and the SGLP action logic during normal plant shutdowns to avoid undesired actuation of the associated equipment. In each case, when three of the four associated measurement channels are below the block setpoint, pressing a manual pushputton will block the actuation signal for that train. If two of the four of the measurement channels increase above the block setpoint, the block will automatically be removed.

The sensor subsystems, including individual channel actuation bistables, is addressed in this LCO. LCO 3.3.4, addresses the The actuation logic subsystems, cofisistifig of the

two out of four manual actuation. and downstream components used to actuate the individual ESF components are addressed in LCO 3.3.4.

Palisades Nuclear Plant B 3.3.3-2 01/20/98 05/30/99

ESF Instrumentation B 3.3.3 BASES BACKGROUND Measurement Channels

{continued)

Measurement channels, consisting of pressure switches, field transmitters, or process sensors and associated instrumentation, provide a measurable electronic signal based upon the physical characteristics of the parameter being measured.

Four identical measurement channels are provided for each parameter used in the generation of trip signals. These are designated Channels A through D. Measurement channels provide input to ESF bi stables within the same ESF channel. In addition, some measurement channels may also be used as inputs to Reactor Protective System {RPS) bistables, and most provide indication in the control room. Measurement channels used as an input te the RPS er ESF are net used fer centrel Functions.

These ESFI sensers shared with the RPS are identified in Table B 3.3.1 1.

When a channel monitoring a parameter indicates an abnormal unsafe condition, the bistable monitoring the parameter in that channel will trip. In the case of RAS and CHP, the sensors are latching auxiliary relays from level and pressure switches, respectively, which do not develop an analog input to separate bistables. Tripping two or more channels ef bistables monitoring the same parameter will actuate de energize both channels of Actuation Logic of the associated ESF equipment.

Three of the four measurement and bistable channels are necessary to meet the redundancy and testability of GDC 21 in Appendix A to 10 CFR 50 {Ref. 2). The fourth channel provides additional flexibility by allowing one channel to be removed from service in a bypass cenditien for maintenance or testing while still maintaining a minimum two-out-of-three ~ogic.

There are, hewever, ne built in previsions for channel bypasses in the ESF design.

Since no single failure will either cause er prevent a protective system actuation and no protective channel feeds a control channel, this arrangement meets the requirements of IEEE Standard 279 7-9-1971 (Ref. 3) .

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ESF Instrumentation B 3.3.3 BASES BACKGROUND Measurement Channels (continued)

The ESF Actuation Functions are generated by comparing a single measurement to a fixed bistable setpoint. The ESF Actuation Functions utilize the following input instrumentation:

Safety Injection Signal (SIS)

The Safety Injection Signal can be generated by any of three inputs: Pressurizer Low Pressure, Containment High Pressure, or Manual Actuation. Manual Actuation is addressed by LCO 3.3.4; Containment High Pressure is discussed below. Four instruments, channels A through D,

monitor Pressurizer Pressure to develop the SIS actuation. Each of these instrument channels has two individually adjustable ESF bistable trip devices, one for the bypass removal circuit (discussed below) and one for SIS. Each ESF bistable trip device actuates two

auxiliary relays, one for each actuation train. The output contacts from these auxiliary relays fonn the logic circuits addressed in LCO 3.3.4. The instrument channels associated with each Pressurizer Low Pressure SIS actuation bistable include the pressure measurement loop, the SIS actuation bistable, and the two auxiliary relays associated with that bistable. The bistables associated with automatic removal of the Pressurizer Low Pressure Bypass are discussed under Function 7.a, below.

Low Steam Generator Pressure Signal (SGLP)

There are two separate Low Steam Generator Pressure signals, one for each steam generator. For each steam generator, four instruments (channels A through D) monitor pressure to develop the SGLP actuation. Each of these instrument channels has two individually adjustable ESF bistable trip devices, one for the bypass removal circuit (discussed below) and one for SGLP. Each Steam SGLP bistable trip device actuates an auxiliary relay.

The output contacts from these auxiliary relays fonn the SGLP logic circuits addressed in LCO 3.3.4. The instrument channels associated with each Steam Generator Low Pressure Signal bistable include the pressure measurement loop, the SGLP actuation bistable, and the auxiliary.relay associated with that bistable. The bistables associated with automatic removal of the SGLP Bypass are discussed under Function 7.a, below.

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ESF Instrumentation B 3.3.3 BASES BACKGROUND Measurement Channels (continued)

Recirculation Actuation Signal (RAS)

There are four Safety lnjecti on Refueling Water (S IRW)

Tank level instruments used to develop the RAS signal.

Each of these instrument channels actuates two auxiliary relays, one for each actuation train. The output contacts from these auxiliary relays form the logic circuits addressed in LCO 3.3.4. The SIRW Tank Low Level instrument channels associated with each RAS actuation bistable include the level instrument and the two auxiliary relays associated with that instrument.

Auxiliary Feedwater Actuation Signal (AFAS)

There are two separate AFAS signals (AFAS chan_nels A and

- B), each one actuated on low level in either steam*

_generator. For each steam generator, four level instruments (channels A through D) monitor level to develop the AFAS actuation signals. The output contacts from the bistables on these level channels form the SGLP logic circuits addre~sed in LCO 3.3.4. The instrument channels associated with each Steam Generator Low Level Signal bistable include the level measurement loop and the Low Level AFAS bistable.

Containment High Pressure Actuation (CHP)

The Containment High Pressure signal is actuated by two sets of four pressure switches, one set for each train.

The output contacts from these pressure switches form the CHP logic circuits addressed in LCO 3.3.4 .

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ESF Instrumentation B 3.3.3 BASES BACKGROUND Measurement Channels (continued)

Containment High Radiation Actuation (CHR)

The Safety Injection Signal can be generated by either of two inputs: High Radiation or Manual Actuation. Manual Actuation is addressed by LCO 3.3.4. Four radiation monitor instruments, channels A through D, monitor containment area radiation level to develop the CHR signal. Each CHR monitor bistable device actuates one auxiliary relay which has contacts in each CHR logic train addressed in LCO 3.3.4. The instrument channels associated with each CHR actuation bistable include the radiation monitor itself and the associated auxiliary relay.

Automatic Bypass Removal Functions

Pressurizer Low Pressure and Steam Generator Low Pressure logic circuits have the capability to be blocked to avoid undesired actuation when pressure is intentionally lowered during plant shutdowns. In each case these bypasses are automatically removed when the measured pressure*exceeds the bypass permissive setpoint. the measurement channels which provide the bypass removal signal are the same channels which provide the actuation signal. Each of these pressure measurement channels has two bistables, one for actuation and one for the bypass removal Function. The pressurizer pressure channels include an auxiliary relay actuated by the bypass removal bistable. The logic circuits for Automatic Bypass Removal Functions are addressed by LCO 3.3.4.

Several measurement instrument channels provide more than one required function. Those sensors shared for RPS and ESF functions are identified in Table B 3.3.1-1. That table provides a listing of those shared channels and the Specifications which they affect .

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ESF Instrumentation B 3.3.3 BASES BACKGROUND Bistable Trip Units (continued)

Where bBistable trip units are used, they receive an analog i np,ut from the measurement device diar-rnel s, compare the analog input to trip setpoints, and provide contact output to the Actuation Logic. They also provide local trip indication and remote annunciation.

There are four channels of bistables, designated A through D, for each ESF Function, one for each measurement channel.

The trip setpoints and Allowable Values used in the bistables are based on the analytical limits stated in Reference 4. The selectioH of these trip setpoiHts is such that adequate protectioH is provided when all sensor and processiHg time delays ar~ takeH iHto account. To allow for calibration tolerances and instrumentation uncertainties, Allowable Values specified in Table 3.3.3 l, in the accompanyiHg LCO, are

coHservatively adjusted with respect to the analytical limits .

The methodology used te validate the trip setpoints is provided iH the plaHt documents. The actual nominal trip setpoint eHtered inte the bistable is normally still more conservative than that specified by the Allowable Value to account fer chaHges in random measurement errors detectable by a CllANNEL FUNCTIONAL TEST aHd CllANNEL CALIBRATION. If the measured setpoint does not exceed the Allowable Value, the bistable is COHSidered OPERABLE.

The Allowable Values are specified for each safety related ESF trip Function which is credited in the safety analysis.

Nominal trip setpoints are specified in the plant procedures.

The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. The methedology used to determine the nominal trip setpoints is also provided irr plant documents.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function.These uncertainties are addressed as described in plant documents. A channel is inoperable if its actual setpoint is not within its Allowable Value .

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ESF Instrumentation B 3.3.3 BASES BACKGROUND Setpoints in accordance with the Allowable Value will ensure (continued} that Safety Limits of Chapter 2.0, "SAFETY LIMITS (SLs)," are not violated during Anticipated Operational Occurrences (AOOs) and that the -consequences of Design Basis Accidents (DBAs) will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or OBA and the equipment functions as designed.

ESF Instrument Channel Bypasses The only ESF instrument channels with built in bypass capability are the Low SG Level AFAS bistables. Those bypasses are effected by a key operated switch, similar to the RPS Trip Channel Bypasses. A bypassed Low SG Level channel AFAS bistable cannot perfonn its specified function and must be considered inoperable.

While there are no other built-in provisions for instrument channel bypasses in the ESF design (bypassing any other channel output requires opening a circuit link, lifting a lead, or using a jumper), this LCO includes requirements for OPERABILITY-of the instrument channels and bistables which provide input to the Automatic Bypass Removal Logic channels required by LCO *3.3.4, "ESF Logic and Manual Initiation."

The Actuation Logic channels for Pressurizer Pressure and Steam Generator Low Pressure, however, have the ability to be manually bypassed when the associated pressure is below the range where automatic protection is required. These actuation logic channel bypasses may be manually initiated when three-out-of-four bypass pennissive bistables indicate below their setpoint. When two-out-of-four of these bistables are ~bove their bypass permissive setpoint, the actuation logic channel bypass is automa ti ca 11 y removed. The bypass permissive bistables use the same four measurement channels as the blocked ESF function for their inputs .

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ESF Instrumentation B 3.3.3 BASES APPLICABLE Many of the analyzed accidents can be detected by one or SAFETY ANALYSES more ESFt Functions. One of the ESFt Functions is the primary actuation signal for that accident. An ESFt Function may be the primary actuation signal for more than one type of accident. An ESFt Function may also be a secondary, or backup, actuation signal for one or more other accidents. Functions not specifically credited in the accident analysis, serve as backups and are part of the NRC approved licensing basis for

the pl ant.

ESFt protective Functions are as follows.

1. Safety Injection Signal (SIS)

The SIS ensures acceptable consequences during Loss of Coolant Accident (LOCA) events, including steam generator tube rupture, and Main Steam Line Breaks (MSLBs) or Feedwater Line Breaks (FWLBs) (inside containment). To provide the required protection, SIS is actuated by a CHP signal, or by two-out-of-four Pressurizer Low Pressure channels decreasing below the setpoint. SIS initiates the following actions:

a. Start HPSI & LPSI pumps;

b. Start component cooling water and service water pumps;

c. Initiate service water valve operations;

d. Initiate component cooling water valve operations;

e. Start containment cooling fans (when coincident with a loss of offsite power); *

f. Enable Containment Spray Pump Start on CHP; and

g. Initiate Safety Injection Valve operations.

Each SIS logic train is also actuated by a contact pair on one of the CHP initiation relays for the associated CHP train .

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ESF Instrumentation B 3.3.3 BASES APPLICABLE Sr. Containment High Pressure Signal (CHP)

SAFETY ANALYSES (continued) The CHP signal closes all containment isolation valves not required for ESF operation and starts containment spray (if SIS enabled), ensuring acceptable consequences during LOCAs, control rod ejection events, MSLBs, or FWLBs (inside containment).

CHP is actuated by two-out-of-four pressure switches for the associated train reaching their setpoints. CHP initiates the following actions:

a. Containment Spray;

b. Safety Injection Signal;

c. Main Feedwater Isolation;

d.

e.

f.

Main Steam Line Isolation; Control Room HVAC Emergency Mode; and Containment Isolation Valve Closure.

6~. Containment High Radiation Signal (CHR)

CHR is actuated by two-out-of-four radiation monitors exceeding their setpoints. CHR initiates the following actions to ensure acceptable consequences following a LOCA or control rod ejection event:

a. Control Room HVAC Emergency Mode;

b. Containment Isolation Valve Closure; and

c. Block automatic starting of ECCS pump room sump pumps.

During refueling operations, separate switch-selectable radiation monitors initiate CHR, as addressed by LCO 3.3.6 .

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ESF Instrumentation B 3.3.3 BASES APPLICABLE 24. Steam Generator Low Pressure Signal (SGLP)

SAFETY ANALYSES (continued) The SGLP ensures acceptable consequences during an MSLB or FWLB by isolating the steam generator if it indicates a low steam generator pressure. The SGLP concurrent with or following a reactor trip, minimizes the rate of heat extraction and subsequent cool down of the PCS during these events.

One SGLP circuit is provided for each SG. Each SGLP circuit is actuated by two-out-of-four pressure channels on the associated SG reaching their setpoint. SGLP initiates the following actions:

a. Close the associated Feedwater Regulating valve and its bypass; and

b. Close both Main Steam Isolation Valves .

35. Recirculation Actuation Signal At the end of the injection phase of a LOCA, the SIRWT will be nearly empty. Continued cooling must be provided by the ECCS to remove decay heat. The source of water for the ECCS pumps is automatically switched to the containment recirculation sump. Switchover from SIRWT to the containment sump must occur before the SIRWT empties to prevent damage to the ECCS pumps and a loss of core cooling capability. For similar reasons, switchover must not occur before there is sufficient water in the containment sump to support pump suction.

Furthermore, early switchover must not occur to ensure sufficient borated water is injected from the SIRWT to ensure the reactor remains shut down in the recirculation mode.- An SIRWT Low Level signal initiates the RAS.

RAS initiates the following actions:

a. Trip LPSI pumps (this trip can be manually bypassed);

b. Switch HPSl'and containment spray pump suction from SIRWT to Containment Sump by opening sump CVs and

c.

closing SIRWT CVs; and Adjust cooling water to component cooling heat exchangers.

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BASES ESF Instrumentation B 3.3.3 APPLICABLE 35. Recirculation Actuation Signal (continued)

SAFETY ANALYSES The RAS signal is actuated by separate sensors from those which provide tank level indication. The allowable range of 21" to 27" above the tank floor corresponds to 1.1% to 3.3% indicated level. Typically the actual setting is near the midpoint of the allowable range.

46. Auxiliary Feedwater Actuation Signal An AFAS initiates feedwater flow to both steam generators if a low level is indicated in either steam generator.

The AFAS maintains a steam generator heat sink during the following events:

MSLB;

FWLB;

7.

LOCA; and Loss of feedwater.

Automatic Bypass Removal Functions The logic circuitry provides automatic removal of the Pressurizer Pressure Low and Steam Generator Pressure Low actuation signal bypasses. There are no assumptions in the safety analyses which assume operation of these automatic bypass removal circuits, and no analyzed events result in conditions where the automatic removal would be required to mitigate the event. The automatic removal circuits are required to assure that logic circuit bypasses will not be overlooked during a plant startup.

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

LCO The LCO requires all channel components necessary to provide an ESF actuation to be OPERABLE.

The Bases for the LCO on ESF Functions are addressed below.

1. Safety Injection Signal (SIS)

This LCO requires four channels of SIS Pressurizer Low Pressure to be OPERABLE in MODES l, 2, and 3.

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BASES ESF Instrumentation B 3.3.3 LCO 1. Safety Injection Signal (SIS) (continued)

The setpoint was chosen so as to be low enough to avoid actuation during plant operating transients, but to be high enough to be quickly actuated by a LOCA or MSLB.

The settings include an uncertainty allowance which is consistent with the settings assumed in the MSLB analysis (which bounds the settings assumed in the LOCA analysis).

Sr. Containment High Pressure Signal (CHP)

This LCO requires four channels of CHP to be OPERABLE for each of the associated ESF trains (left and right) in MODES 1, 2, 3 and 4.

The setpoint was chosen so as to be high enough to avoid actuation by containment temperature or atmospheric

6~.

pressure changes, but low enough to be quickly actuated by a LOCA or a MSLB in the containment.

Containment High Radiation Signal (CHR)

This LCO requires four channels of CHR to be OPERABLE in MODES 1, 2, 3, and 4.

The setpoint is based on the maximum primary coolant leakage t6 the containment atmosphere allowed by LCO 3.4.13 and 16the maximum activity allowed by LCO 3.4.16. N .concentration reaches equilibrium in containment atmosphere due to its short half-life, but other activity was assumed to build up. At the end of a 24 hou~ leakage period the dose rate i~ approximately 20 R/h as seen by the area monitors. A large leak could cause the area dose rate to quickly exceed the 20 R/h setting and initiate CHR .

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ESF Instrumentation B 3.3.3 BASES LCO 24. Steam Generator Low Pressure Signal (SGLP) I (continued)

This LCO requires four channels of Steam Generator Low Pressure Instrumentation for each SG to be OPERABLE in MODES 1, 2, and 3. However, as indicated in Table 3.3.3-1, Note (a), the SGLP Function is not required to be OPERABLE in MODES 2 or 3 if all Main Steam Isolation Valves (MSIVs) are closed and deactivated and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves.

Th~ setpoint was chosen to be low enough to avoid actuation during plant operati6n, but be close enough to full power operating pressure to be actuated quickly in the event of a MSLB. The setting includes an uncertainty allowance which is consistent with the setting used* in the Reference 4 analysis .

Each SGLP logic is made up of output contacts from four pressure bistables from the associated SG. When the logic circuit is satisfied, two relays are energized to actuate steam and feedwater line isolation.

This LCO applies to failures in the four sensor subsystems, including sensors, bistables, and associated equipment. Failures in the actuation subsystems, incltlein~ the maHtlal bypass switches, are considered Actuation Logic failures and are addressed in LCO 3.3.4.

35. Recirculation Actuation Signal (RAS)

This LCO requires four channels of SIRWT Low Level to be OPERABLE in MODES 1, 2, and 3.

The setpoint was chosen to provide adequate water in the containment sump for HPSI pump net positive suction head following an accident, but.prevent the pumps from running dry during the switchover .

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ESF Instrumentation B 3.3.3 BASES LCO 35. Recirculation Actuation Signal (RAS) (continued)

The upper limit on the Allowable Value for this trip is set low enough to ensure RAS does not initiate before sufficient water is transferred to the containment sump.

Premature recirculation could impair the reactivity control Function of safety injection by limiting the amount of boron injection. Premature recirculation could also damage or disable the recirculation system if recirculation begins before the sump has enough water.

The lower limit on the SIRWT Low Level trip Allowable Value is high enough to transfer suction to the containment sump prior to emptying the SIRWT.

45. Auxiliarv Feedwater Actuation Signal (AFAS)

The AFAS logic actuates AFW to a SG on a SG Low Level in that SG.

The Allowable Value was chosen to assure that AFW flow would be initiated while the SG could still act as a heat sink and steam source, and to assure that a reactor trip would not occur on low level without the actuation of AFW~

This LCO requires four channels for each steam generator of Steam Generator Low Level to be OPERABLE in MODES 1, 2, and 3.

7. Automatic Bypass Removal The automatic bypass removal logic removes the bypasses which are used during plant shutdown periods, for Pressurizer Low Pressure and Steam Generator Low Pressure actuation signals.

The setpoints were chosen to be above the setpoint for the associated actuation signal, but well below the normal operating pressures.

\

This LCO requires four channels of Pressurizer Low Pressure bypass removal and four channels for each steam generator of Steam Generator Low Pressure bypass removal; to be OPERABLE in MODES 1, 2, and 3.

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ESF Instrumentation B 3.3.3 BASES APPLICABILITY All ESF Functions are required to be OPERABLE in MODES 1, 2, and 3. In addition, Containment High Pressure and Containment High Radiation are required to be operable in MODE 4.

In MODES 1, 2, and 3 there is sufficient energy in the primary and secondary systems to warrant automatic ESF System responses to:

Close the main steam isolation valves to preclude a positive reactivity addition and containment overpressure;

Actuate AFW to preclude the loss of the steam generators as a heat sink (in the event the normal feedwater system is not available);

Actuate ESF systems to prevent or limit the release of fission product radioactivity to the environment by isolating containment and limiting the containment pressure from exceeding the containment design pressure during a design basis LOCA or MSLB; and

Actuate ESF systems to ensure sufficient borated inventory to permit adequate core cooling and reactivity control during a design basis LOCA or MSLB accident.

The CHP and CHR Functions are -a+s-t> required to be OPERABLE in MODE 4 to limit leakage of radioactive material from containment and limit operator exposure during and following a OBA.

The SGLP Function is not required to be OPERABLE in MODES 2 and 3, if all MSIVs are closed and deactivated and all MFRVs and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves, since the SGLP Function is not required to perform any safety functions under these conditi ans.

In lower MODES, automatic actuation of ESF Functions is not required, because adequate time is available for plant .

operators to evaluate plant conditions and respond by manually operating the ESF components, if required. LCD 3.3.6 addresses automatic Refueling CHR isolation during CORE ALTERATIONS or during movement of irradiated fuel .

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BASES ESF Instrumentation 8 3.3.3 APPLICABILITY. In MODES 5 and 6, ESFAS initiated systems are either (continued) reconfigured or disabled for shutdown cooling operation.

Accidents in these MODES are slow to develop and would be mitigated by manual operation of individual components.

ACTIONS The mest eemmeR eause ef ehaRRel iReperability is eutri~ht failure er arift ef the bistable er precess meaule suffieieRt te exceea the teleraRee allewea by the plaRt specific setpeiRt aRalysis.

The most common causes of channel inoperability are outright failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel

- failing to the "safe" condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards).

Typically, the drift is small and results in a delay of actuation rather than a total loss of function. Determination of setpoint drift is generally made during the performance of a CHANNEL FUNCTIONAL TEST when the process instrument is set up for adjustment to bring it to within specification. If the actual trip setpoint is not within the Allowable Value in Table 3.3.3-1, the channel is inoperable and the appropriate Condition(s) are entered.

In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value in Table 3.3.3-1, or the sensor, instrument loop, signal processing electronics, or ESFt bistable is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the plant must enter the Condition statement for the particular protection Function affected .

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ESF Instrumentation B 3.3.3 BASES ACTIONS When the number of inoperable channels in a trip Function (continued) exceeds those specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered if applicable in the current MODE of operation.

A Note has been added to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for ~~ch Function:in Table 3.3.3-1. Com~letion Times for the inoper~ble channel of a Function will be tracked separately.

Condition A applies to the failure of a single bistable or associated instrumentation channel of one or more input parameters in each ESFt Function except the SIRWT Low Level RAS I *

Function. Since the bistable and associated instrument channel combine to perform the actuation function, the Condition,is also appropriate if both the bistable and associated instrument channel are inoperable.

ESFt coincidence logic is normally two-out-of-four. If one ESFt channel is inoperable, startup or power operation is allowed to continue as long as action is taken to restore the design level of redundancy.

If one ESFt channel is inoperable, startup or power operation is allowed to continue, providing _the inoperable channel actuation bistable is placed in trip within 7 days. The provision of four trip channels allows one channel to be failed inoperable in a non-trip condition or optioHally bypassed up to the 7 day Completion Time allotted to place the channel in trip, although except for AFAS, there are HO iHstalled desigH provisioHs for this bypass fuHctioH. Operating with one failed channel in a non-trip condition or Bypassed (removed from service) during operations, places the ESF Actuation Logic in a two-out-of-three coincidence logic.

If the failed channel cannot be restored to OPERABLE status in 7 days, the associated bistable is placed in a tripped condition. This places the function in a one-out-of-three configuration .

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I ESF Instrumentation B 3.3.3 BASES ACTIONS A.1 (continued)

In this configuration, common cause failure of the dependent channel cannot prevent ESF actuation. The 7 day Completion Time is based upon operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.

Condition A is modified by a Note which indicates it is not applicable to the SIRWT Low Level Function.

B.1 and B.2 Condition B applies to the failure of two channels in any of the ESFt Functions except the SIRWT Low Level RAS Function.

With two inoperable channels, one channel actuation device must be placed in trip within the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time. Eight

hours is allowed for this action since it must be accomplished by a circuit modification, or by removing power from a circuit component. With one channel of protective instrumentation bypassed, or failed in a non trip condition inoperable, the ESF Actuation Logic Function is in two-out-of-three logic, but with another channel failed inoperable the ESFt may be operating with a two-out-of-two logic. This is outside the assumptions made in the analyses and should be corrected. To correct the problem, the second channel is placed in trip. This places the ESFt in a one-out-of-two logic. If any of the other OPERABLE channels receives a trip signal, ESFt actuation will occur.

One of the failed channels must be restored to OPERABLE status within 7 days, and the provisions of Condition A still applied to the remaining inoperable channel. Therefore, the channel that is still inoperable after completion of Required Action B.2 must be placed in trip if more than 7 days has elapsed since the channel's initial failure.

Condition B is modified by a Note which indicates that it is not applicable to the SIRWT Low Level Function. The Required Action is also modified by a Note stating that LCO 3.0.4 is not applicable. The Note was added to allow the changing of MODES even though two channels are inoperable, with one channel tripped. MODE change~ in this configuration are allowed, to permit maintenance and testing on one of the inoperable channels. In this configuration, the protection system is in a one-out-of-two logic, and the probability of a common cause failure affecting both of the OPERABLE channels during the 7 days permitted is remote.

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ESF Instrumentation B 3.3.3 BASES ACTIONS C.1 and C.2 (continued)

Condition C applies to one RAS SIRWT Low Level channel inoperable. The SIRWT low level circuitry is arranged in a

"l-out-of-2 taken twice" logic rather than the more frequently used 2-out-of-4 logic. Therefore, Required Action C.1 differs from other ESF functions. With a bypassed SIRWT low level channel, an additional failure might disable automatic RAS, but would not initiate a premature RAS. With a tripped channel, an additional failure could cause a premature RAS, but would not disable the automatic RAS.

Since considerable time is available after initiation of SIS until RAS is required and there is quite a tolerance on the time when RAS must be initiated, and since a premature RAS could damage tt++ the ESF pumps, it is preferable to bypass an inoperable channel and risk loss of automatic RAS than to trip

. a channel and risk a, premature RAS .

The Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months allowed is reasonable because the Required Action involves a circuit modification.

Required Action C.2 requires that the inoperable channel be re~airee afte restored to OPERABLE status within 7 days. The Completion Time is reasonable based upon operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.

D.1 and D.2 If the Required Actions and associated Completion Times of Condition A, B, or C are not met for Functions 1, 2, 3, 4, or 71, 4, 5, or 6, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to 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 3.3.3-20 01/20/98 05/30/99

ESF Instrumentation B 3.3.3 BASES ACTIONS E.1 and E.2 (continued)

If the Required Actions and associated Completion Times of Condition A, B, or C are not met for Functions 5r or 6~, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE The SRs for any particular ESFt Function are found in the REQUIREMENTS SRs column of Table 3.3.3-1 for that Function. Most functions are subject to CHANNEL CHECK, CHANNEL FUNCTIUNAL TEST, and CHANNEL CALIBRATION .

While Palisades is not committed to performing all testing discussed in ANSI/IEEE Standard 338-1977, CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, AND CHANNEL CALIBRATIONS are performed in accordance with the guidance of ANSI/IEEE Standard 338-1977, which is endorsed by Regulatory Guide 1.118.

I I

SR 3.3.3.1 A CHANNEL CHECK is performed once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months on each ESFt input channel which is provided with an indicator to provide a qualitative assurance that the channel is working properly and that its readings are within limits. A CHANNEL CHECK is not performed on the CHP and SIRWT Low Level channels because they have no associated control room indicator.

Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious . . CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to

operate properly between each CHANNEL CALIBRATION.

Palisades Nuclear Plant 8 3.3.3-21 01/20/98 05/30/99

BASES ESF Instrumentation B 3.3.3 SURVEILLANCE SR 3.3.3.1 (continued)

REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties. -including indication and readability. If a channel is outside the criteria. it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when Surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction.

Offscale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.

The Frequency of about once every shift is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period is extremely low, the CHANNEL CHECK

minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of CHANNEL OPERABILITY during normal operational use of displays associated with the LCO required channels.

SR 3.3.3.2 A CHANNEL FUNCTIONAL TEST is performed every 92 days to ensure the entire channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

This test is required to be performed each 92 days on ESFT input channels provided with on-line testing capability. It is not required for the SIRWT Low Level channels since they have no built in test capability. The CHANNEL FUNCTIONAL TEST for SIRWT Low Level channels is performed each 18 months as part of the required CHANNEL CALIBRATION .

The CHANNEL FUNCTIONAL TEST tests the individual se"sor subsystems channels using an analog test input to each bistable.

Palisades Nuclear Plant B 3.3.3-22 01/20/98 05/30/99

ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.2 (continued)

REQUIREMENTS A test signal is superimpesee en the input in ene channel at a time te verify that the bistable trips within the specifiee telerance areune the setpeint. Any setpoint adjustment shall be consistent with the assumptions of the current p+attt-specific setpoint analysis.

The as feune ane as left values must alse be recereee ane reviewee fer censistency with the assumptiens ef the surveillance interval extensien analysis. The requirements fer this review are eutlinee in Reference 5.

The Frequency .of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Reference 5). '

SR 3.3.3.3 CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. CHANNEL CALIBRATIONS must be performed consistent with the p+attt-specific setpoint analysis.

The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the extension analysis. The requirements for this review are outlined in Reference 5.

The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis .

Palisades Nuclear Plant B 3.3.3-23 01/20/98 05/30/99

ESF Instrumentation B 3.3.3 BASES REFERENCES 1. FSAR, Chapter 7

2. 10 CFR 50, Appendix A

3. IEEE Standard 279-1971

4. FSAR, Chapter 14

5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989 Palisades Nuclear Plant B 3.3.3-24 01/20/98 05/30/99

ESF Logic and Manual Initiation B 3.3.4

B 3.3 INSTRUMENTATION B 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation BASES BACKGROUND The ESF Instrumentation (ESFI) initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System (PCS) pressure boundary and to mitigate accidents.

The ESFI cefltaifls devices afld circuitry tfta-t-generates the following signals listed below when the monitored variables reach levels that are indicative of conditions requiring protective action.

Alse listed are the inputs to each ESF Actuation Signal are also listed.

1. Safety Injection Signal (SIS);

a. Containment High Pressure (CHP)

b. Pressurizer Low Pressure 5r. Containment High Pressure Signal (CHP);

a. Containment High Pressure - Left Train

b. Containment High Pressure - Right Train 6~. Containment High Radiation Signal (CHR)

.a. Cont~inment High Radiation

24. Steam Generator Low Pressure Signal (SGLP)

a. Steam Generator -f56t-A Low Pressure

b. Steam Generator B Low Pressure J5. Recirculation Actuation Signal (RAS);

a. Safety Injection Refueling Water Tank (SIRWT) Low Level Palisades Nuclear Plant B 3.3.4-1 01/20/98 05/30/99

ESF Logic and Manual Ini ti ati on

BASES BACKGROUND 46. Auxiliary Feedwater Actuation Signal {.Ii.FAS)

B 3.3.4 (continued)

a. Steam Generator A Low Level

b. Steam Generator B Low Level In the above list of actuation signals, the CHP and RAS are derived from pressure and level switches, respectively.

Equipment actuated by each of the above signals is identified in the FSAR, Chapter 7 (Ref. 1).

Th~ ESF circuitry, with the exception of RAS, employs two-out-of-four logtc. Four independent measurement channels are provided for each function used to generate ESF actuation signals. When any two channels of the same function reach their setpoint, actuating relays are efiergized w~ich, ifi tHrfi, initiate the protective actions. Two separate and redundant trains of actuating relays, each powered from separate power supplies, are utilized. These separate relay trains operate redundant trains of ESF equipment. The actuation relays are considered part of the actuation logic addressed by this LCO.

RAS logic consists of output contacts of the relays actuated by the SIRWT Low Level switches arranged in a "one-out-of-two taken twice" logic. The contacts are arranged so that at least one low level -signal powered from each station battery is required to initiate RAS. Loss of a single battery, therefore, cannot either cause or prevent RAS initiation.

The sensor subsystem, including individual channel bistables, is addressed in LCO 3.3.3, "Engineered Safety Features (ESF)

Instrumentation." This LCO addresses the actuation subsyste1111 iofisistifig of the two OHt of foHr manual actuation, and downstream components used to actuate the individual ESF functions, as defined in the following section.

ESE Logic Each of the six ESF actuation actHatifig signals in Table 3.3.4-1 operates two trains of actuating relays. Each train is capable of initiating the ESF equipment load groHps to meet the minimum requirements to provide all functions necessary to operate the system associated with the plant's capability to cope with abnormal events .

Palisades Nuclear Plant B 3.3.4-2 01/20/98 05/30/99

BASES ESF Logic and Manual Initiation B 3.

3.4 BACKGROUND

ESF Logic (continued)

The SGLP logic circuitry includes bypass provisions such that the SGLP automatic actuation Function may be bypassed (blocked) blocked if three-out-of-four Steam GeneratQr (SG) pressure channels are below a bypass block permissive setpoint.

Similarly, the SIS automatic actuation on Pressurizer Low Pressure may be bypassed blocked when three-out-of-four channels' are below a permissive setpoint. This bypassing actuation blocking is performed when the ESF Functions these inputs are no longer required for protection. These bypasses actuation, blocks are enabled manually when the enabling permissive conditions are satisfied in three of the four sensor subsystem channels.

The operating bypass circuitry employs four bistable channels in the sensor subsystems, sensing pressurizer pressure (for the SIS) and SG pressure (for the SGLP). These bistables provide contact output to the three out of four logic in the two

actuation logic channels. When the logic is satisfied, manual bypassing is permitted. There are two manual bypass actuation controls for each Function, one per train.

All operating bypasses actuation blocks are automatically removed when enabling bypass conditions are no longer satisfied. If an SIS or SGLP automatic actuation channel is blocked, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable.

Failure of the bistable circuitry used to iHiti ate the block permissive is considered a measurement channel or bistable channel failure and is addressed by LCO 3.3.3. Failure in the legie used to effect two out of fe~r bypass removal er failure of the manual bypass enable circuitry to remo*9'e U1e bypass is addressed by this LCO.

Testing of a major portion of the ESF circuits*is accomplished while the plant is at power. More extensive sequencerand load testing may be done with the reactor shut down. The test circuits are designed to test the redundant circuits separately such that the correct operation of each circuit may be verified by either equipment operation or by sequence lights.

\

j Palisades Nuclear Plant B 3.3.4-3 01/20/98 05/30/99

BASES ESF Logic and Manual. Initiation B 3.

3.4 BACKGROUND

Manual Initiation (continued)

Manual ESF initiation capability is provided to permit the operator to manually actuate an ESF System when necessary.

Two control room mounted manual actuation switches are provided for the SIS and CllR actuation, one for each train. CllP and RAS may be initiated using individual component controls. In the case of SIS, each Each SIS manual actuation switch affects one actuation channel, which actuates one train of ES-FSIS equipment ..

The~e ar~ no single manual controls provided to actuate CHP, however, CHP may be manually initiated using individual component controls.

Two control room mounted manual actuation switches are provided for CHR actuation, each switch affects both actuation channels, which actuates both CHR trains .

There are no single manual controls provided to actuate SGLP, however, SGLP may be manually initiated using individual component controls. Main Steam Isolation Valves (MSI'Js) are provided with two closure switches in the control room. Either switch closes both MSI'Js. Other SGLP actuated components must be manually operated using individual component controls.

There are no single manual controls provided to actuate RAS, however, RAS may be manually Jnitiated using individual component controls.

Manual actuation of AFW may be accomplished through pushbutton actuation of each AFAS channel or by use of individual pump and valve controls. Each automatic AFAS actuation channel starts the associated AFW pump{s) pumps in their starting sequence (if P-8A fails to start, a P-8C start signal is generated, and if P-8C also fails to start, a P-8B start signal is generated) and opens the associated flow control valves.

RAS does not possess separate manual switches. To actuate a RAS manually, it is necessary to actuate the individual components from the control room or use the "test" switches, each of which actuates one train. Two channels of RAS manual actuation are shown in Table 3.3.4 1. Each channel consists of either the individual 'component manual or "test" switches for one train .

Palisades Nuclear Plant B 3.3.4-4 01/20/98 05/30/99

BASES ESF Logic and Manual Initiation B 3.3.4 APPLICABLE Many of the analyzed accidents can be detected by one or SAFETY ANALYSES more ESF Functions. One of the ESF Functions is the primary actuation signal for that accident. An ESF Function may be the primary actuation signal for more than one type of accident. '

An ESF Function may also be a secondary, or backup, actuation signal for one or more other accidents. Functions such as Manual Initiation, not specifically credited in the accident analysis, serve as backups to Functions and are part of the NRC staff approved licensing basis for the plant.

The manual initiation is not required by the accident analysis.

The ESF logic must function in all situations where the ESF function is required (as discussed in the Bases for LCO 3.3.3).

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

LCO . The LCO requires that all components necessary to provide an

ESF actuation be OPERABLE .

The Bases for the LCD on ESF automatic actuation Functions are addressed in LCD 3.3.3. Those associated with the Manual Initiation or Actuation Logic are addressed below.

ESF Logic and Manual Initiation Functions are required to be OPERABLE in MODES 1, 2, and 3, or in MODES 1, 2, 3, and 4, as appropriate, when the associated automatic initiation channels addressed by LCO 3.3.3 are required.

1. Safety Injection Signal (SIS)

SIS is actuated by manual initi~tion, by a CHP signal, or by two-out-of-four Pressurizer Low Pressure channels decreasing below the setpoint.

Each Manual Initiation channel consists of one pushbutton which directly starts the SIS actuation logic for the associated train.

Each SIS logic train is also actuated by a contact pair on one of the CHP initiation relays for the associated CHP train.

\

a. Manual Initiationf-rt-tt

This LCD requires two ~hannels of SIS Manual Initiation to be OPERABLE.

Palisades Nuclear Plant B 3.3.4-5 01/20/98 05/30/99

ESF Logic and Manual Initiation B 3.3.4 BASES LCO b. Actuation Logic (continued)

This LCO requires two channels of SIS Actuation Logic to be OPERABLE.

Failures in the actuation subsystems, iRclueiRg the maRual bypass switches, are ActuatioR Logic failures aRe are addressed in this LCO.

c. CHP Logic Trains The CHP initiation relay (5P-x) input to the SIS logic is considered part of the SIS logic. Two channels, one per SIS train, must therefore be OPERABLE.

d. Automatic Bypass Removal

\._

This LCO requires fettr two channels of the automatic bypass permissive removal logic for SIS Pressurizer Low Pressure to be OPERABLE in MODES 1, 2, and 3. If an SIS automatic actuation channel is bypassed, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable.

The Pressurizer Low Pressure logic train for each SIS train can be bypassed when three-out-of-four channels indicate below 1700 psia. This bypass prevents undesired actuation of SIS during a normal plant cooldown. The bypass signal is automatically removed when two-out-of-four channels exceed the setpoint, in accordance with the bypass philosophy of removing bypasses when the enabling conditions are no longer satisfied.

The bypass permissive chaRRels coRsist of four seRsor subsystems aRe trw*o actuati OR subsystems. This LCO applies to Failures iR the four seRsor subsystems, iRelueiRg seRsors, bistables, aRe associatee equipmeRt.

Failures iR the actuatioR subsystems, iRclueiRg the maRual bypass switches, are coRsieeree ActuatioR Lo~ie failures aRe are aeeressee iR LCO 3.3.4. This LCO also applies to the bypass removal feature. If the bypass eRable FuRctioR is failee so as to preveRt eRteriRg a bypass eoReitioR, operatioR may coRtiRue.

The bypass permissive is set low enough so as not to be enabled during normal plant operation, but high enough to allow bypassing prior to reaching the trip setpoint.

Palisades Nuclear Plant B 3.3.4-6 01/20/98 05/30/99

ESF Logic and Manual Initiation B 3.3.4

BASES LCO (continued) 5r. Containment High Pressure Signal (CHP)

a. Manual Initiation This LCO requires the manual controls netessary to actuate those valves and components actuated by an automatic CHP to be OPERABLE.

ba. Actuation Logic This LCO requires two channels of CHP Actuation Logic to be OPERABLE.

6~. Containment High Radiation Signal (CHR)

a. Manual Initiation This LCO requires two channels of CHR Manual Initiation to be OPERABLE. Pushbuttons are available for manual actuation of each CHR logic train .

b. Actuation Logic This LCO requires two channels of CHR Actuation Logic to be OPERABLE.

24. Steam Generator Low Pressure Signal (SGLP)

a. Manual Initiation This LCO requires two channels of SGLP Manual Initiation to be OPERABLE. There is no manual control which actuates the SGLP logic circuits. The actuated components must be individually actuated using control room manual controls.

b. Actuation Logic This LCO requires two channels of SGLP Actuation Logic to be OPERABLE, one for each SG .

Palisades Nuclear Plant B 3.3.4-7 01/20/98 05/30/99

__J

ESF Logic and Manual Initiation B 3.3.4 BASES LCO c. Automatic Bypass Removal (Continued)

This LCO requires two channels, one for each SG, of the SGLP automatic bypass removal logic to be OPERABLE in MODES 1, 2, and 3. If an SGLP automatic actuation channel is bypassed, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable.

The SGLP from each SG may be bypassed when three of the four SG pressure three-out-of-four channels indicate below 565 psia. This bypass prevents undesired actuation during a normal plant cooldown.

The bypass signal is automatically removed when steam pressure exceecls two-out-of-four channels exceed the setpoint, in accordance with the bypass philosophy of removing bypasses when the enabling conditions are no longer satisfied.

Each SGLP logic is macle up of output coRtacts from

*four pressure bistables from the associatecl SG .

WheR the logic circuit is satisfiecl, two relays are eRergizecl to actuate steam aRcl feeclwater liRe isolatioR. A similar logic circuit is proviclecl for each block circuit.

This LCO also applies to the bypass removal feature.

If the bypass eRable FuRctioR is failecl so as to preveRt eRteriRg a bypass coRclitioR, operatioR may CORtiRue.

The bypass permissive is set low enough so as not to be enabled during normal plant operation, but high enough to allow bypassing prior to reaching the trip setpoint.

35. Recirculation Actuation Signal (RAS)

a. Manual Initiation This LCO requires two channels of RAS Manual Initiation to be OPERABLE. There is no manual control of which actuates the RAS logic circuits.

The actuated components must be individually actuated using manual controls .

Palisades Nuclear Plant B 3.3.4-8 01/20/98 05/30/99 I

_J

ESF Logic and Manual Initiation B 3.3.4 BASES LCO b. Actuation Logic (continued)

This LCO requires two channels of RAS Actuation Logic to be OPERABLE.

46. Auxiliary Feedwater Actuation Signal (AFAS)

a. Manual Initiation This LCO requires two channels of AFAS Manual Initiation to be OPERABLE. Each train of AFAS may be manually initiated with either of two sets of controls. Only one set of manual controls is required to be OPERABLE for each AFW train. One set of c*ontrols are the pushbuttons provided to actuate

each train on the C-11 panel; the other set of controls are those manual controls provided on C-01 for each AFW pump and flow control valve .

b. Actuation Logic This LCO requires two channels of AFAS Actuation Logic to be OPERABLE.

APPLICABILITY At+-ESF Functions are required to be OPERABLE in MODES 1. 2.

and 3 or MODES 1, 2, 3, and 4 as specified in Table 3.3.4-1-as-apprepriate. In these MODES 1, 2. and 3, there is sufficient energy in the primary and secondary systems to warrant automatic ESF System responses to:

Close the MSIVs to preclude a positive reactivity addition and containment overpressure;

Actuate AFW to preclude the loss of the steam generators as a heat sink (in the event the normal feedwater system is not available);

Actuate ESF systems to prevent or limit the release of fission product radioactivity, to the environment by isolating containment and limiting the containment pressure from exceeding the containment design pressure

during a design basis LOCA or MSLB; and Palisades Nuclear Plant 8 3.3.4-9 01/20/98 05/30/99

ESF Logic and Manual Initiation B 3.3.4 BASES APPLICABILITY

Actuate ESF systems to ensure sufficient borated (continued) inventory to pennit adequate core cooling and reactivity control during a design basis LOCA or MSLB accident.

The CHP and CHR Functions are also required to be OPERABLE in MODE 4 to limit leakage of radioactive material from containment and limit operator exposure during and following a DBA.

The SGLP Function is not required to be OPERABLE in MODES 2 and 3, if all MSIVs are closed and deactivated and all MFRVs and MF~V bypass valves are either closed and deactivated or isolated by closed manual valves, since the SGLP Function is not required to perform any safety function under these cond it i ans.

In lower MODES 5 and 6, automatic actuation of ESF Functions is not required, because adequate ti.me is available for plant operators to evaluate plant conditions anij respond by manually operating the ESF components if required. In these MODES --*- itftd--6-, ESF initiated systems are either reconfigured or disabled for shutdown cooling operation. Accidents in these MODES are slow to develop and would be mitigated by manual operation of individual components.

ACTIONS When the number of inoperable channels in a trip Function exceeds those specified in a-ny related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered, if applicable in the current MODE of operation.

A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for e*ach Function in Table 3.3.4-1 in the LCO. Completion Times for the inoperable channel of a Function will be tracked separately.

Condition A applies to one Manual Initiation, Bypass Removal, or Actuation Logic, or Bypass Removal channel inoperable.

The channel must be restored to OPERABLE status to restore redundancy of the ESF Function. The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time is commensurate with the importance of avoiding the vulnerability of a single failure in the only remaining OPERABLE channel.

Palisades Nuclear Plant B 3.3.4-10 01/20/98 05/30/99

ESF Logic and Manual Initiation 8 3 .3. 4

BASES ACTIONS (continued)

B.1 and B.2 If two Manual Initiation, Bypass Removal, or Actuation Logic channels inoperable for Functions 1, 2, 3, or 4, 5, or 6, or if the Required Action and associated Completion Time of Condition A cannot be met for Function l, 2, 3, or 4, 5, or 6, which are Applicable iH MODES 1, 2, aRd 3, t~e reactor must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to 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.

C.1 and C.2 Condition C is entered when one or more Functions have two Manual Initiation, Bypass Removal, or Actuation Logic channels inoperable for Functions 5r or 6~. and when the Required Action

and associated Completion Time of Condition A are not met for Functions 2 or 35 or 6. If Required Action A.1 cannot be met within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach*

the required plant conditions from full power conditions in an orderly manner .and without challenging plant systems.

SURVEILLANCE Refer to Table 3.3.4 1 to determiRe which SR shall be REQUIREMENTS performed for each FuRctioR.

SR 3.3.4.1 A functional test of each SIS actuation channel must be performed each 92 days. This test is to be performed using the installed control room test switches and test circuits for both "with standby power" and "without offsite standby power". When testing the "with standby power~ circuits, proper operation of the SIS-X relays must be verified; when testing the "without standby power" circuits, proper operation of the OBA sequencer and the associated logic circuit must be verified. The test circuits are designed to block those SIS functions, such as

injection of concentrated boric acid, which would interfere with plant operation.

Palisades Nuclear Plant B 3 .3. 4-11 01/20/98 05/30/99

l ESF Logic and Manual Initiation B 3.3.4 BASES SURVEILLANCE SR 3.3.4.1 (continued)

REQUIREMENTS The Frequency of 92 days is based on the reliability analysis presented in topical report CEN 327, "RPS/ESFAS Extended Test Interval Evaluation (Ref. 2)a feature of the initial Palisades license.

SR 3.3.4.2 A CHANNEL FUNCTIONAL TEST of each AFAS Actuation Logic Channel is performed every 92 days to ensure the entire channel will perfonn its intended function when needed. Sensor subsystem tests are addressed in LCO 3.3.3. A successful test of the I required contact(s) of a channel relay may be perfonned by the I verification of the change of state of a single contact of the I relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL I TEST of a relay. This is acceptable because all of the other I required contacts of the relay are verified by other Technical I Specifications and non-Technical Specifications tests at least I

once per refueling interval with applicable extensions.

Sensor subsystem Instrumentation channel tests are addressed in LCO 3.3.3.

fft-i-5-SR 3.3.4.2 addresses Actuation Logic tests of the AFAS

.I using the installed test circuits.

This SR is modified by a Note which states that Actuation Logic tests include operation of initiation relays.

The Frequency of 92 days for SR 3.3.4.2 is in agreement with the conclusions of ~based OH the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 2).

SR 3.3.4.3 A CHANNEL FUNCTIONAL TEST is performed on the manual ESF actuation circuitry channels, bypass removal channels, and Actuation Logic channels for certain ESF Functions, providing actuation of the Function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required ~ontacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Palisades Nuclear Plant B 3.3.4-12 01/20/98 05/30/99

ESF Logic and Manual Initiation B 3.3.4 BASES SURVEILLANCE SR 3.3.4.3 (continued)

REQUIREMENTS This Surveillance verifies that the trip pusR butto"s of tRe Manual Initiation Functions are capable of ope"i"g co"tacts i" tRe Actuatio" Logic as desig"ed, providi"g Ma"ual I"itiatio" of tRe Fu"ctio"OPERABLE. This Surveillance also verifies that the entire channel of the Manual Actuation Logic will perfonn its intended function when needed.

The 18 month Frequency is based on the need to perfonn this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were perfonned with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when perfonned at a Frequency of once every 18 months.

REFERENCES 1. FSAR, Chapter 7

2. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989 Palisades Nuclear Plant 8 3.3.4-13 01/20/98 05/30/99

DG - UV Start B 3.3.5 B 3.3 INSTRUMENTATION B 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)

BASES BACKGROUND The DGs provide a source of emergency power when offsite power is either unavailable or insufficiently stable to allow safe plant operation. Undervoltage protection will generate a UV Start in the event a Loss of Voltage or Degraded Voltage condition occurs. There are two UV Start Functions for each 2.4 kV vital bus.

Undervoltage protection and load shedding features for safety-related buses at the 2,400 V and lower voltage levels are designed in accordance with IO CFR 50, Appendix A, General Design Criterion I7 (Ref. I) and the following features:

1. Two levels of automatic undervoltage protection from loss or degradation of offsite power sources are provided.

The first level (loss of voltage) provides nonnal loss of voltage protection. The second level of protection (degraded voltage) has voltage and time delay set points selected for automatic trip of the offsite sources to protect safety-related equipment from sustained degraded voltage conditions at all bus voltage levels.

Coincidence logic is provided to preclude spurious_ trips.

2. The undervoltage protection system automatically prevents

.. load shedding of the safety-related buses when the emergency generators are supplying power to the safeguards loads. *

\

3. Control circuits for shedding of Class IE and non-Class IE loads during a Loss of Coolant Accident (LOCA) themselves are Class IE or are separated electrically from the Class IE portions .

Palisades Nuclear Plant B 3.3.5-I 01/20/98 05/30/99

DG - UV Start B 3.3.5 BASES BACKGROUND Description (continued)

Each 2,400 V Bus (lC and lD) is equipped with two levels of undervoltage protection relays (Ref. 2). The first level (Loss of Voltage Function) relays 127-1 and 127-2 are set at approximately 77% of rated voltage with an inverse time relay.

E-aeft One of these relays measures voltage on a++ each of the three phases. They iHttl protect~ against sudden loss of voltage as sensed on the corresponding bus using a three-out-of-three coincidence logic. The actuation of ft the associated auxiliary relays will trip 4-t-s- the associated bus respective incoming btt5".

circuit breakers, start its associated DG, initiate bus load shedding, and activate annunciators in the control room. The DG circuit breaker is closed automatically upon establishment of satisfactory voltage and frequency by the use of associated t:ttttiervoltage pretectien sensing relay 127D-1 or 127D-2.

The second level of undervoltage protection (Degraded Voltage Function) relays 127-7 and 127-8 are set at approximately 93~% I

of rated voltage, with eaeh one relay monitoring a++ each of the three phases en its respective bus. These relays protect against sustained degraded voltage conditions on the corresponding bus using a three-out-of-three coincidence logic.

These relays have a built-in 0.65&:-5 second time delay, after which he-tftthe associated DG~ receives a start signal and activate annunciators in the control room are actuated. If a I

I bus undervoltage exists after an additional six seconds, the respective associated bus incoming btt5" circuit breakers will be tripped and a bus load shed will be initiated.

Trip Setpoints .and Allewable Values The trip setpoints and Allewable Values are based on the analytical limits presented in References 3 and 4, and justified in Reference 5. The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account. To allow for calibration tolerances, instrumentation uncertainties, and instrument drift, Allewable Values setpoints I specified in SR 3.3.5.f2 are conservatively adjusted with -I respect to the analytical limits .. A detailed analysis of the degraded voltage protection is provided in References 3 and 4.

Palisades Nuclear Plant B 3.3.5-2 01/20/98 05/30/99

DG - UV Start B 3.3.5 BASES BACKGROUND Trip Setpoints aMd Allowable Values (continued) j The specified s£etpoints iM accordaMce with the Allowable Values will ensure that the consequences of accidents will be acceptable, providing the plant is operated from within the LCOs at the onset of the accident and the equipment functions as designed.

APPLICABLE The DG - UV Start is required for Engineered Safety Features SAFETY ANALYSES (ESF) systems to function in any accident with a loss of offsite power. Its design basis is that of the ESF Systems.

Accident analyses credit the loading of the DG based on a l-0ss of offsite power during a LOCA. The diesel loading has been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power. This delay time includes contributions from the

DG start, DG loading, and Safety Injection System component actuation.

The required channels of UV Start, in conjunction with the ESF systems powered from the DGs, provide plant protection in the event of any of the analyzed accidents discussed in

Reference 6, in which a loss of offsite power is assumed. UV Start channels are required to meet the redundancy and testability requirements of GDC 21 in 10 CFR 50, Appendix A (Ref. 1).

The delay times assumed in the safety analysis for the ESF equipment include the 10 second DG start delay and.the appropriate sequencing delay, if applicable. The response times for ESFAS actuated equipment include the appropriate DG loading and sequencing delay.

The DG - UV Start channels satisfy Criterion 3 of 10 CFR 50.36(c)(2)

Palisades Nuclear Plant B 3.3.5-3 01/20/98 05/30/99

DG - UV Start B 3.3.5 BASES LCO The LCO for the DG - UV Start requires that threeette channels per bus of each UV Start instrumentation Function be OPERABLE when the associated DG is required to be OPERABLE. The UV Start supports safety systems associated with ESF actuation.

Loss of DG UV Start Function could result in the delay of safety system initiation when required. This could lead to unacceptable consequences during accidents. During the loss of offsite power, which is an anticipated operational occurrence, the DG powers the motor driven auxiliary feedw*ater pumps.

Failure of these pumps to start would leave only the one turbine driven pump as well as an increased potential for a loss of decay heat removal through the secondary system.

Only Allowable Values are specified for each Function in the LCO. Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure that the setpoint measured by. CllANNEL CALIBRATIONS does not exceed the Allowable Value if the bistable is performing as required.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within the Allowable Value, is acceptable, provided that operation and testing are consistent with the assumptions of the plant specific setpoint calculation. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.

The Bases for the Allowable Values and trip setpoints are as foll OWS*:

The voltage trip setpoint is set low enough such that spurious trips of the offsite source due to operation of the undervoltage relays are not expected for any combination of plant loads and nonnal grid voltages~

This setpoint at the 2,400 V bus and reflected down to the 480 V buses has been verified through an analysis to be greater than the minimum allowable motor voltage (90% of nominal

voltage). Motors are the most limiting equipment in the system. MCC contactor pickup and drop-out voltage is also adequate at the setpoint values. The analysis ensures that the distribution system is capable of starting and operating all safety-related equipment within the equipment voltage rating at the allowed source voltages. The power distribution system model used in the analysis has been verified by actual testing (Refs. 5 and 7)

Palisades Nuclear Plant B 3.3.5-4 01/20/98 05/30/99

DG - UV Start B 3.3.5 BASES LCO The time delays involved will not cause any thermal damage (continued) as the setpoints are within voltage ranges for sustained operation. They are long enough to preclude trip of the offsite source caused by the starting of large motors and yet do not exceed the time limits of ESF actuation assumed in FSAR Chapter 14 (Ref. 6) and validated by Reference 8.

Calibration of the undervoltage relays verify that the time delay is sufficient to avoid spurious trips.

APPLICABILITY The DG - UV Start actuation Function is required to be OPERABLE whenever the required associated DG must be is required to be OPERABLE per LCO 3.8.1, "AC Sources - Operating," or LCO 3.8.2, "AC Sources - Shutdown," so that it can perform its function on a loss of power or degraded power to the vital bus .

ACTIONS A DG - UV Start channel is inoperable when it does not satisfy the OPERABILITY criteria for the channel 1 s Function.

In the event a channel s trip setpoint is found nonconservative 1

with respect to the Specified SetpointAllewable Value, or the channel is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the LCO Condition entered. The required channels are specified on a per DG basis.

Condition A applies if one or more of the three phase UV sensors or relay logic is inoperable for one or more Functions (Degraded Voltage or Loss of Voltage) per DG bus.

If the ehaRRel eaRRot be restored te OPERABLE status, tThe affected DG must be declared inoperable and the appropriate Condition(s) entered. Because of the three-out-of-three logic in both the Loss of Voltage and Degraded Voltage Functions,

eembiRed with the abseRee ef readily available ehaRRel bypass eapability, the appropriatemest expeditious means of addressing*

channel failure is delaring the DGehaRRel inoperable, and effecting repair in a manner consistent with other DG failures

Palisades Nuclear Plant B 3.3.5-5 01/20/98 05/30/99

DG - UV Start

BASES ACTIONS A.1 (continued)

B 3.3.5 Required Action A.1 ensures that Required Actions for the affected DG inoperabilities are initiated. Depending upon plant MODE. the actions specified in LCO 3.8.1 or LCO 3.8.2 as 1

applicable, are required immediately.

suRVEILLANCE SR 3.3.5.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST i.s perfonned on each UV Start logic channel every 18 months to ensure that the logic channel will perfonn its intended function when needed. The Under Voltage sensing relays are tested by SR 3.3.5.2. A successful test of the required contact(s) of a channel relay may be perfonned by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other

.Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

The Frequency of 18 months is based on the plant necessary to perform the test.

co~ditions The fellewifig SR applies te each DG UV Start Fufietiefi.

SR 3.3.5.2t SR 3.3.5.lis the perfermafiee ef aA CHANNEL CALIBRATION every perfonned each 18 months. ne CllANNEL CALIBRATION verifies the accuracy of each component within the instrument channel. This includes calibration of the undervoltage relays and demonstrates that the equipment falls within the specified operating characteristics defined by the manufacturer.

The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests.

CHANNEL CALIBRATIONS must be perfonned consistent with the setpoint analysis .

The Frequency of 18 months is based on the plant conditions necessary to perform the test.

Palisades Nuclear Plant B 3.3.5-6 01/20/98 05/30/99

DG - UV Start B 3.3.5 BASES REFERENCES 1. 10 CFR 50, Appendix A GDCs 17 and 21

2. FSAR, Section 8.6

3. CPCo Analysis EA-ELEC-VOLT-033

4. CPCo Analysis EA-ELEC-VOLT-034

5. CPCo Analysis EA-ELEC-VOLT-17

6. FSAR, Chapter 14

7. CPCo Analysis EA-ELEC-VOLT-13 8 . . CPCo Analysis A-NL-92-111 Palisades Nuclear Plant B 3.3.5-7 01/20/98 05/30/99

Refueling CHR Instrumentation B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Refueling Containment High Radiation (CHR) Instrumentation BASES BACKGROUND This LCO addresses Refueling CHR actuation. When the Refueling I CHR Monitors are enabled, a CHR actuation may be automatically I initiated by a signal from either of the Refueling CHR monitors I or manually by actuation of either of the control room "CHR I.

Manual Initiate" pushbuttons (pushing either Manual Initiate I pushbutton will actuate both trains of CHR). A CHR signal I initiates the following actions: I I

a. Control Room HVAC Emergency Mode; I I

b. Containment Isolation Valve Closure; and I I

c. Block automatic starting of Engineered Safeguards pump I room sump pumps . I

The Refueling CHR signal provides automatic containment isolation valve closure during refueling operations, using two radiation monitors located in the refueling area of the containment (elevation 649 ft). The monitors are part of the plant area monitoring system and employ one-out-of-two logic for isolation. During normal operation these monitors will not initiate an isolation a CHR signal. A switch is provided so that isolation CHR actuation can be initiated by these monitors during refueling only.

Each monitor actuates one train of CHR logic when containment radiation exceeds the setpoint. Two separate enabling keylock switches, one per train, enable the Refueling CHR input to the CHR logic when switched to the "Refueling" Mode. Each Refueling CHR channel, associated keylock switch, and initiation circuit input to the CHR logic thus forms a one-out-of-one logic input to its associated CHR actuation logic train.

The Refueling CHR isolation instrumentation is separate from the CHR .instrumentation addressed in LCO 3.3.3, "ESF Instrumentation." However, the Refueling CHR Instrumentation does operate the same CHR actuation relays as the two-out-of-four CHR logic addressed in LCO 3.3.4. This LCO is not included in LCOs 3.3.3 and 3.3.4 because of the differences in APPLICABILITY and the,single channel nature of the Refueling CHR input. The Refueling CHR signal performs the automatic

containment isolation valve closure Function during refueling operations req~ired by LCO 3.9.3, "Containment Penetrations."

Palisades Nuclear Plant B 3.3.6-1 01/20/98 05/30/99

Refueling CHR Instrumentation B 3.3.6 BASES BACKGROUND The Refueling CHR Instrumentation provides protection from (continued) release of radioactive contamination gases and particulates from +tt the containment in the event a fuel assembly should be severely damaged during handling.

The Refueling CHR Instrumentation will detect any abnonnal radiation levels in the containment refueling area and will initiate purge valve closure to limit the release of radioactivity to the environment. The containment purge supply and exhaust valves are closed on a CHR signal when a high radiation level in containment i~ detected.

The Refueling CHR Instrumentation includes two independent.

redundant actuation subsystems. as described above. Reference 1 describes the Refueling CHR circuitry.

Trfp Setpoints-

No required setpoint is specified because these instruments are I not assumed to function by any of the safety analyses.

Typically, the instruments are set at about 25 mR/hr above expected background for planned operations (including movement of the reactor vessel head or internals).

Setpoints in ~ccordance with the allowed range will ensure th~

consequences ef Design Basis Accidents 1v'i 11 ae acceptaal e.

. .I

I I

I APPLICABLE The Refueling CHR Instrumentation isolates containment in SAFETY ANALYSES the event that area radiation exceeds an established level following a fuel handling accident~ This ensures the radioactive materials are not released directly to the environment and significantly reduces the offsite doses from those calculated without considering by the safety analyses, I which do not credit containment isolation (Ref. 2). Either I*

way, i.e., with or without containment isolation, the offside offsite doses remain within the guidelines of 10 CFR 100.

The Refueling CHR Instrumentation is not required by the fuel handling accident analyses to maintain offsite doses within_ the guidelines of 10 CFR 100, but operating experience indicates that containment isolation provides significant reduction of the resulting offsite'doses. Therefore, the Refueling\ CHR Instrumentation satisfies the requirements of Criterion 4

of 10 CFR 50.3o(c)(2) .

Palisades Nuclear Plant B 3.3.6-2 01/20/98 05/30/99

Refueling CHR Instrumentation B 3.3.6 BASES LCO The LCO for the Refueling CHR Instrumentation requires that two channels of refueling CHR instrumentation and two channels of CHR manual initiationQoth channels be OPERABLE, including the logic components necessary to initiate Refueling CHR Isolation.

Operation with a trip setpoint less conservative than the nominal trip setpoi n,t, btJt *within its all mtee rarige, is acceptable. The CHR setpoint ts chosen to be high enough to avoid inadvertent actuation in the event of nbnnal background radiation fluctuations during fuel handling and movement of the reactor internals, but low enough to alann and isolate the containment in the event of a Design Basis fuel handling accident.

APPLICABILITY In MODE 5 or 6, the Refueling CHR isolation of containment isolation valves is not nonnally required to be OPERABLE.

However, during CORE ALTERATIONS or during movement of irradiated fuel within containment, there is the possibility of a fuel handling accident requiring containment isolation on high radiation in containment. Accordingly, the Refueling CHR Instrumentation must be OPERABLE during CORE ALTERATIONS and when moving any irradiated fuel in containment.

In MODES 1, 2, 3 and 4, both the Containment High Pressure (CHP) and CHR signals provide containment isolation as discussed in the Bases for LCO 3.3.3 and LCO 3.3.4.

ACTIONS~~~~--11~n-t~hMe~e~ve~n~t:..-na-P-1ch~a~n~n~e~1~

1 s~t~r1H*p~s~e~tp~o~i~n~t--+<is~f~o~tJ~ne~n~a~n~co~n~s~e~r~va~t~i~v~e with respect to the allowee range, or the associatee instrtJmentation channel is fotJne inoperable, then the ReftJeling CllR ftJHCti OH provi eee by that channel shotJl e be eecl aree inoperable ane LCO Coneition A enteree.

A.1. A.2.1. and A.2.2 Condition A applies to the failure of one Refueling CHR monitor channel, one CHR Manual Initiate channel, or one of each. The Required Action allows either initiation of a CHR signal by placing the inoperable channel in trip (which accomplishes the safety function of th~ inoperable channel), or suspension of CORE ALTERATIONS and movement of irradiated fuel assemblies

within containment (which places the plant in a condition where the LCO does not apply). The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is warranted because one additional channel of each Function remains operable during that period.

Palisades Nuclear Plant B 3.3.6-3 01/20/98 05/30/99

Refueling CHR Instrumentation B 3.3.6 BASES ACTIONS A.l. A.2.1. and A.2.2 (continued)

The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe*

position .

. B.1 and B.2 Condition B applies when either no automatic Refueling CHR or no Manual CHR (or neither) is available. The Required Action is to immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. This places the plant in a condition where the LCO does not apply. The Completion Time is warranted on the basis that at least one containment isolation Function is completely lost.

The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position

A.I afld A.2 Coflditiofl A applies to the failure of Ofle or Both Refueliflg CllR ehaflflels. The Required Aetiofl is to immediately suspefld CORE ALTERATIONS afld movemeflt of irradiated fuel assemBlies withifl cofltaiflmeflt. This places the plafit ifl a coflditiofl where the LCO does flot apply. The Completiofl Time is warraflted of! the Basis that cof!taiflmeflt isolatiofl capaBility is lost.

The suspeflsiofl of CORE ALTERATIONS afld fuel movemeflt shall flOt preclude eompletiofl of movemeflt of a compofleflt to a safe pos iti Ofl.

SURVEILLANCE SR 3.3.6.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value .

Palisades Nuclear Plant B 3.3.6-4 01/20/98 05/30/99

Refueling CHR Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.1 (continued)

REQUIREMENTS Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or ef semething even mere serious actual differing radiation levels at the two detector locations. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff, based en a cembinatien ef the channel instrument uncertainties, including indicatien and readability. If a channel is eutside the criteria, it may be an indicatien that the transmitter er the signal precessing equipment has drifted eutside its limits.

The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure.

Since the probability of two random failures in redundant channels in any 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels.

SR 3.3.6.2 A CHANNEL FUNCTIONAL TEST is performed on each Refueling CHR channel to ensure the entire channel will perform its iDtended function. A successful test of the required contact(s) of a I channel relay may be performed by the verification of the I change of state of a single contact of the relay. This I clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a I relay. This is acceptable because all of the other required *I contacts of the relay are verified by other Technical I Specifications and non-Technical Specifications tests at least I once per refueling interval with applicable extensions. Any I setpeint adjustment must be censistent with 10 CFR 100 requirements.

The Frequency of 31 days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given

Function in any 31 day interval is a rare event.

Palisades Nuclear Plant B 3.3.6-5 01/20/98 05/30/99

_J

BASES Refueling CHR Instrumentation B 3.3.6 SURVEILLANCE SR 3.3.6.3 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each CHR Manual Initiation channel to ensure it will perform its intended function.

The Frequency of 18 months is based on plant operating experience with regard to channel OPERABILITY, and is consistent with the testing of other manually actuated functions.

SR 3.3.6.4 A CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies

. that the channel responds t~ a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between

successive calibrations to ensure that the channel remains operational between successive tests. CllANNEL CALIBRATIONS mHst be performed consistent with the setpoint determination.

The Frequency is based upon the assumption of an 18 month calibration interval in the setpoint determination.

REFERENCES 1. FSAR, Section 7.3

2. FSAR, Section~14.19 Palisades Nuclear Plant B 3.3.6-6 01/20/98 05/30/99

PAM Instrumentation

B 3.3 INSTRUMENTATION B 3.3.7 Post Accident Monitoring (PAM) Instrumentation B 3.3.7 BASES BACKGROUND The primary purpose of the Post Accident Monitoring (PAM) instrumentation is to display plant variables that provide infonnation required by the control room operators during accident situations. This infonnation provides the necessary support- for the operator to take the manual actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety Functions for Design Basis Events.

The OPERABILITY of the PAM instrumentation ensures that there is sufficient infonnation available on selected plant parameters to monitor and assess plant status and behavior following an accident.

The availability of PAM instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be detennined.

These essential instruments are identified in the FSAR Appendix 7C (Ref. 1) and address the recommendations of Regulatory Guide 1.97 (Ref. 2), ~s required by Supplement 1 to NUREG-0737, 11 TMI Action Items 11 (Ref. 3).

Type A variables are included in this LCO because they provide the primary infonnation required to pennit the control room operator to take specific manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions for Design Basis Accidents (DBAs).

Category I variables are the key variables deemed risk significant because they are needed to:

-* Detenni ne whether other systems important to safety are perfonning their intended functions;

Provide infonnation to the operators that will enable them to detennine the potential for causing a gross breach of the barriers to radioactivity release; and

Provide infonnation regarding the release of radioactive materials to allow for early indication of the need to

initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.

Palisades Nuclear Plant B 3.3.7-1 01/20/98 05/30/99

BASES PAM Instrumentation B 3.

3.7 BACKGROUND

These key variables are identified in Reference 1 by plant (continued) specific Regulatory Guide 1.97 analyses (Ref. 1). This analysis identified the plant specific Type A and Category 1 variables and provided justification for deviating from the NRC proposed list of Category I variables.

The specific instrument Functions listed in Table 3.3.7-1 are discussed in the LCO Bases.

APPLICABLE The PAM instrumentation ensures the OPERABILITY of SAFETY ANALYSES Regulatory Guide 1.97 Type A variables, so that the control room operating staff can:

Perfonn the diagnosis specified in the emergency operating procedures. These variables are restricted to preplanned actions for the primary success path of DBAs; and

Take the specified, preplanned, manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions.

The PAM instrumentation also ensures OPERABILITY of Category I, non-Type A variables. This ensures the control room operating staff can:

Detennine whether systems important to safety are perfonning their intended functions;

Detennine the potential for caµsing a gross breach of the barriers to radioactivity release;

Detennine if a gross breach of a barrier has occurred; and

Initiate action necessary to protect the public as well as to obtain an estimate of the magnitude of any impending threat.

PAM instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97'meets Criterion 3 of 10 CFR 50.36(c)(2) .

Palisades Nuclear Plant B 3.3.7-2 01/20/98 05/30/99

PAM Instrumentation

BASES B 3.3.7 APPLICABLE Category I, non-Type A PAM instruments are retained in the SAFETY ANALYSES Specification because they are intended to assist operators (continued) in minimizing the consequences of accidents. Therefore, these Category I variables are important in reducing public risk.

LCO LCO 3.3.7 requires at least two OPERABLE channels for all but one Function to ensure no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following that accident.

Furthermore, provision of at least two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information.

The exception to the two channel requirement is Containment Isolation Valve Position. In this case, the important infonnation is the status of the containment penetrations. The LCO requires one position indicator for each active containment isolation valve. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of the passive valve or via system boundary status. If a normally active containment isolation valve is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE.

Listed below are discussions of the specified instrument Functions listed in Table 3.3.7-1. Component identifiers of the sensors, indicators, power supplies, displays, and recorders in each instrument loop are found in Reference 1.

Palisades Nuclear Plant B 3.3.7-3 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES LCO 1, 2. Primary Coolant System (PCS) Hot and Cold Leg

{continued) Temperature (wide range)

PCS wide range Hot and Cold Leg Temperatures are Type B, Category 1 variables provided for verification of core cooling and long term surveillance.

Reactor outlet temperature inputs to the PAM are provided by two wide range resistance elements and associated transmitters (one in each loop). The channels provide indication over a range of 50°F to 700°F.

3. Wide Range Neutron Flux Wide Range Neutron Flux indication is a Type B, Category 1 variable, and is provided to verify reactor shutdown .

4. Containment Floor Water Level (wide range)

Wide range Containment Floor Water Level is a Type B, Category 1 variable, and is provided for verification and long term surveillance of PCS integrity.

5. Subcooled Margin Monitor The Subcooled Margin Monitor (SMM) is a Type A, Category 1 variable used to identify conditions

which require tripping of the primary coolant pumps and throttling of safety injection flows. Each SMM channel uses a number of PCS pressure and temperature inputs to determine the degree of PCS subcooling or superheat .

Palisades Nuclear Plant B 3.3.7-4 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES LCO 6. Pressurizer Level (Wide Range)

(continued)

Pressurizer Level is a Type A, Category 1 variable, and is used to determine whether to terminate Safety Injection (SI). if still in progress, or to reinitiate SI if it has been stopped. Knowledge of pressurizer water level is also used to verify the plant conditions necessary to establish natural circulation in the PCS and to verify that the plant is maintained in a safe shutdown condition.

7. Containment Hydrogen Monitors Containment Hydrogen Monitors are provided to detect high hydrogen concentration conditions (a Type A~

Category 1 variable) that represent a potential for containment breach and are used to determine when to place the hydrogen recombiners in operation. This variable is also important in verifying the adequacy of mitigating actions.

8. Condensate Storage Tank (CST) Level CST Level is a Type D, Category 1 variable, and is provided to ensure water supply for AFW. The CST provides the eftstlred safety grade water supply for the AFW System. Inventory is monitored by a O to 100% level indication. CST Level is displayed on a control room indicator. In addition, a control room annunciator alarms on low level.

The CST is the initial source of water for the AFW System. However, as the CST is depleted, manual operator action is necessary to replenish the CST .

Palisades Nuclear Plant B 3.3.7-5 01/20/98 05/30/99

PAM Instrumentation 8 3.3.7 BASES LCO 9. Primary Coolant System Pressure (wide range)

(continued)

PCS wide range pressure is a Type A, Category 1 variable provided for verification of core cooling and PCS integrity long term surveillance.

Wide range PCS loop pressure is measured by pressure transmitters with a span of O psia to 3000 psig.

Redundant monitoring capability is provided by two channels of instrumentation. Control room indications are provided on C12 and C02.

10. Containment Pressure (wide range)

Wide range Containment Pressure is a Type C, Category 1 variable, and is provided for verification of PCS and containment OPERABILITY. It is also an input to decisions for initiating

11, 12.

containment spray .

Steam Generator Water Level (wide range)

Wide range Steam Generator Water Level is a Type A; Category 1 variable, and is provided to monitor operation of decay heat removal via the steam generators. The steam generator level instrumentation covers a span extending from the tube sheet to the steam separators, with an indicated range of -140% to +150%. Redundant monitoring capability is provided by two channels of instrumentation for each SG.

Operator action for maintenance of heat removal is based on the control room indication of Steam Generator Water Level. The indication is used during a SG tube rupture to determine which SG has the ruptured tube. It is also used to determine when to initiate once through cooling on low water level .

Palisades Nuclear Plant B 3.3.7-6 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES LCO 13, 14. SG Pressure (continued)

Steam Generator Pressure is a Type A, Category 1 variable used in accident identification, including Loss of Coolant, and Steam Line Break. Redundant monitoring capability is provided by two channels of instrumentation for each SG. *

15. Containment Isolation Valve Position Containment Isolation Valve (CIV) Position is a Type B, Category 1 variable, and is provided for verification of containment OPERABILITY.

CIV position is provided for verification of containment integrity. In the case of CIV position, the important 1

information is the isolation status of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active CIV in a containment penetration flow path. This is sufficient to redundantly verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status. 1f a penetration flow path is isolated, position indication for the CIV(s) in the associated penetration flow path is not needed to determine status. Therefore, as indicated in Note (a) the position indication for valves in an isolated penetration flow path is not required to be OPERABLE.

16, 17, 18, 19. Core Exit Temperature Core Exit.Temperature is a Type c, Category 1 variable, and is provided for verification and long term surveillance of core cooling.

Each Required Core Exit Thermocouple (CET) channel consists of a single environmentally qualified thermocouple.

The design of the Incore Instrumentation System includes a Type K (chromel alumel) thermocouple within each of the

t&-incore.instrument detector assemblies .

Palisades Nuclear Plant B 3.3.7-7 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES LCO 16, 17, 18, 19. Core Exit Temperature (continued)

The junction of each then11ocouple is located above the core exit, inside the incore detector assembly guide tube, that supports and shields the incore instrument detector assembly string from flow forces in the outlet plenum region. These core,exit then11ocouples monitor the temperature of the reactor coolant as it exits the fuel assemblies.

The core exit then11ocouples have a usable temperature range from 32°F to 2300°F, although accuracy is reduced at temperatures above 1800°F.

20. Reactor Vessel Water Level Reactor Vessel Water Level is monitored by the Reactor Vessel Level Monitoring System (RVLMS) and is a Type B, Category 1 variable provided for verification and long ten11 surveillance of core cooling.

The RVLMS provides a direct measurement of the collapsed liquid level above the fuel alignment plate. The collapsed level represents the amount of liquid mass that is in the reactor vessel above the core. Measurement of the collapsed water level is selected because it is a direct indication of the water inventory. The collapsed level is obtained over the*same temperature and pressure range as the saturation measurements, thereby encompassing all operating and accident conditions where it must function. Also, it functions during the recovery interval. Therefore, it is designed to survive the high steam temperature that may occur during the preceding core recovery interval .

Palisades Nuclear Plant 8 3.3.7-8 01/20/98 05/30/99

BASES PAM Instrumentation B 3.3.7 LCO 20. Reactor Vessel Water Level (continued)

The level range extends from the top of the vessel down to the top of the fuel alignment plate. A total of eight Heated Junction Thennocouple (HJTC) pairs are employed in each of the two RVLMS channels. Each pair consists of a heated junction TC and an unheated junction TC. The differential temperature at each HJTC pair provides discrete indication of uncovery at the HJTC pair location. This indication is displayed using LEDs in the control room. This provides the operator with adequate indication to track the progression of the accident and to detect the consequences of its mitigating actions or the functionality of automatic equipment.

A RVLMS channel consists of eight sensors in a probe. A channel is OPERABLE if four or more sensors, two or more of the upper four and two or more of the lower four, are OPERABLE .

21. Containment Area Radiation (high range)

High range Containment Area Radiation is a Type E, Category 1 variable, and .is provided to monitor for the potential of significant radiation releases and to provide release assessment for use by operators in detennining the need to invoke site emergency plans.

At least t~o channels are required to be OPERABLE for all but one Function. Two OPERABLE channels ensure that no single failure, within either the PAM instrumentation or its auxiliary supporting features or power sources (concurrent with the failures that are a condition of or result from a specific accident), prevents the operators from being presented the information necessary for them to determine the safety status of the plant and to bring the plant to and maintain it in a safe condition following that accident.

In Table 3.3.7 1 the exception to the two channel requirement is Containment Isolation Valve Position.

For steam generator related variables. the required information is individual steam generator level and pressure. In these cases two channels are required to be OPERABLE for each steam generator to redundantly provide the necessary information.

Palisades Nuclear Plant B 3.3.7-9 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES LCO In the ease ef Centainment Iselatien Valve Pesitien, the (eentin1:1ed) impertant infermatien is the stat1:1s ef the eentainment penetratiens. The LCO req1:1ires ene pesition indieator for eaeh aetive eentainment isolation valve. This is suffieient to red1:1ndantly verify the isolatien stat1:1s of each isolable penetration either via indicated stat1:1s of the active valve and prior knewled~e of the passive valve or via system bo1:1ndary stat1:1s. If a normally active containment i sol ati on *tal *1e is known to be closed and deactivated, position indication is not needed to determine stat1:1s. Therefore, the position indication for valves in this state is not req1:1ired to be OPERABLE .

APPLICABILITY . The PAM instrumentation LCD is applicable in MODES 1, 2, and 3.

These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1, 2, and 3. In MODES-4, 5,, and 6, plant conditions are such that the likelihood of an event

ACTIONS occurring that would require PAM instrumentation is low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES.

Note 1 has been added in the ACTIONS to exclude the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS, even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to monitor an accident using alternate instruments and methods, and the low probability of an event requiring these instruments.

Note 2 has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.7-1. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function, starting from the time the Condition was entered for that Function .

Palisades Nuclear Plant B 3.3.7-10 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES ACTIONS . A....l (continued)

When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

This Required Action specifies initiation of actions in accordance with Specification 5.6.6, which requires a written report to be submitted to the Nuclear Regulatory Commission.

This report discusses the re~ults of the root cause evaluation of the inoperability and identifies proposed restorative Required Actions. This Required Action is appropriate in lieu

( of a shutdown requirement, given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs.

C.1 When one or more Functions have two required channels inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrumentation operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an*

accident occur.

Condition C is modified by a Note which indicates it is not

applicable to hydrogen monitor channels .

Palisades Nuclear Plant B 3.3.7-11 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES ACTIONS (continued)

Condition D applies when two hydrogen monitor channels are inoperable. Required Action D.1 requires restoring one hydrogen monitor channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable based on the backup capability of the Post Accident Sampling System to monitor the hydrogen concentration for evaluation of core damage and to prov*ide information for operator decisions. Also, it is unlikely that a LOCA (which would cause core damage) would occur during this time.

When two required hydregen menitor channels are ineperable, Required Actien D.1 requires one channel to be restored to OPERABLE status. This Required Action restores the monitoring capability of the hydregen monitor. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the relatively low probability of an event requiring hydrogen monitoring and the availability of alternative means to obtain the required information.

Continuous operation with two required channels inoperable is not acceptable because alternate indications are not available .

Ll

.)

This Required Action directs entry into the appropriate Condition referenced in Table 3.3.7-1. The applicable Condition referenced in the Table is Function dependent. Each time Required Action C.l or D.1 is.not met, and the associated Completion Time has expired, Condition E is entered for that channel and provides for transfer to the appropriate subsequent Condition.

F.1 and F.2 If the Required Action and associated Completion Time of Condition C or D are not met, and Table 3.3.7-1 directs entry into Condition F, the plant must be brought to a MODE in which the requirements of this LCO do not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 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 3.3.7-12 01/20/98 05/30/99

BASES PAM Instrumentation B 3.3.7 ACTIONS G.1 (continued)

Alternate means of monitoring Reactor Vessel Water Level and Containment Area Radiation have been developed and tested.

These alternate means may be temporarily installed if the nonnal PAM channel cannot be restored to OPERABLE status within the allotted time. If these alternate means are used, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the nonnal PAM channels.

.SURVEILLANCE A Note at the beginning of the Surveillance Requirements REQUIREMENTS specifies that the following SRs apply to each PAM

instrumentation Function in Table 3.3.7-1 .

SR 3.3.7.1 Perfonnance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is nonnally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION

Palisades Nuclear Plant B 3.3.7-13 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.1 (continued)

REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction.

Off scale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.

As indicated in the SR, a CHANNEL CHECK is only required for those channels which are normally energized.

The Frequency of 31 day~ is based upon plant operating .

experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel during normal operational use of the displays associated with this LC0 s required channels.

1 SR. 3.3.7.2 A CHANNEL CALIBRATION is performed every 18 months or approximately every refueling. CHANNEL CALIBRATION is typically a complete check of the instrument channel including the sensor. Therefore, this SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal.

Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes. The Surveillance verifies the channel responds to the measured parameter within the necessary range and accuracy.

For the core exit thermocouples, a CHANNEL CALIBRATION is performed by substituting a known voltage for the thermocouple .

Palisades Nuclear Plant B 3.3.7-14 01/20/98 05/30/99

PAM Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.2 (continued)

REQUIREMENTS The Frequency is based upon operating experience and consistency with the typical industry refueling cycle and is justified by an 18 month calibration interval for the detennination of the magnitude of equipment drift.

REFERENCES 1. FSAR, Appendix 7C, 11 Regulatory Guide 1.97 Instrumentation 11

2. Regulatory Guide 1.97

3. NUREG-0737, Supplement 1 Palisades Nuclear Plant B 3.3.7-15 01/20/98 05/30/99

Alternate Shutdown System B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Alternate Shutdown System BASES BACKGROUND The Alternate Shutdown System provides the control room operator with sufficient instrumentation and controls to maintain the plant in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible. A safe shutdown condition is defined as MODE 3.

With the plant in MODE 3, the Auxiliary Feedwater (AFW) System and the steam generator safety valves or the steam generator atmospheric dump valves can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the AFW System and the ability to borate the Primary Coolant System (PCS) from outside the control room allow extended operation in MODE 3.

In order to ensure use of sufficient components of the AFW System and sufficient process information to permit reactor MODE 3 control in the event a fire damages equipment and circuitry of the main feedwater system or the AFW System in the control room. cable spreading.room. Engineered Safeguards Auxiliary Panel C-33 room, or the corridor between Switchgear Room 1-C and the charging pump rooms, auxiliary Hot Shutdown Control Panels (C-150/C-150A) have been provided and located in the southwest electrical penetration room. These panels are comprised of two enclosures, the main enclosure C-150 and an auxiliary enclosure C-150A. The description below combines these two enclosures into one entity "Panel C-150."

From this panel, control of the AFW flow control valves and control of AFW turbine steam supply Valve B can be enabled.

Indication of AFW flow from the steam driven AFW pump to both Steam Generators (SGs), water level of both SGs, and pressurizer level are enabled by transfer. In addition.

primary coolant pressure (pressurizer pressure) is displayed by a primary sensor dedicated to this use. Transfer of the above-mentioned systems is annunciated in the control room.

See FSAR Section 7.4 (Ref. 1) for operation via Panel C-150.

The equipment controls that are required are listed in Table 3.3.8-lthe LCO sectioR of this basis .

Palisades Nuclear Plant B 3.3.8-1 01/20/98 05/30/99

Alternate Shutdown System

BASES BACKGROUND B 3 .3 .8

Switches, which transfer control or instrument functions *

(continued) from the control room to the auxiliary shutdown control panel, alann in the control room when the devices in the alternate hot shutdown panel are enabled.

APPLICABLE The Alternate Shutdown System is required to provide SAFETY ANALYSES equipment at appropriate locations outside the control room with a capability to maintain the plant in a safe condition in MODE 3.

The criteria governing the design and the specific system requirements of the Alternate Shutdown System are located in 10 CFR 50, Appendix A, GDC 19, and Appendix R (Ref. 2).

The Alternate Shutdown System has been identified as an important contributor to the reduction of plant risk to accidents and, therefore, satisfies the requirements of Criterion 4 of 10 CFR 50.36(c)(2) .

LCO The Alternate Shutdown System LCO provides the requirements for the OPERABILITY of one channel of the instrumentation and controls necessary to maintain the plant in MODE 3 from a location other than the control room. The instrumentation and controls required are listed in Table 3.3.8-1 in the accompanying LCO.

Equipment controls that are required by the alternative dedicated method of maintaining MODE 3 are as follows:

\

1. AFW flow control valves (CV-0727 and CV-0749); and

2. Turbine-driven AFW pump.

Instrumentation systems displayed on the Auxiliary Hot Shutdown Control Pan~l are:

1. Source range flux monitor;

2. AFW flow (HIC-0727 and HIC-0749C);

3. Pressurizer pres~ure;

4. Pressurizer level;

5. SG level and pressure; Palisades Nuclear Plant B 3.3.8-2 01/20/98 05/30/99

Alternate Shutdown System 8 3.3.8 BASES LCO 6. Primary coolant temperatures (hot and cold legs);

(continued)

7. Turbine-driven AFW pump low-suction pressure warning 1i ght; and

8. SIRW tank level.

A Function of an Alternate Shutdown System is OPERABLE if all instrument and control channels needed to support the remote shutdown Functions are OPERABLE.

The Alternate Shutdown System instrumentation and control circuits covered by this LCD do not need to be energized to be considered OPERABLE. This LCD is intended to ensure that the instrument and control circuits will be OPERABLE if plant conditions require that the Alternate Shutdown System be placed in operation.

Table 3.3.8-1 Indication Channel 1, Source Range Nuclear Instrumentation, uses the same detector and preamplifier as the control room channel. Optical isolation is provided between the control room and AHSDP (Alternate Hot Shut Down Panel) portions of the circuit. When the control switches are changed to the "AHSDP" position, the detector and preamplifier is isolated from its normal power supply and connected into the AHSDP power supply.

Table 3.3.8-1 Indication Channels 2 and 12 are provided with their own pressure and level transmitter. The associated circuitry is energized when the AHSDP is energized.

The other Table 3.3.8-1 Indication Channels 3 through 11, aftd t-4--in Table 3.3.8-1 use a transmitter which also serves normal control room instrumentation. When the control switches are changed to the "AHSDP" (Alternate Hot Shut Down Panel) position, the transmitter is isolated from its normal power supply and circuitry, and connected into the C-150 or C-150A panel circuit; control for AFW flow control valves CV-0727 and CV-0749 is also transferred to C-150. The transfer switches are alarmed in the control room.

Pressurizer pressure ifidieator ehafifiel 2 is provided with its owfi pressure trafismitter. Its eireuitry is eftergized whefi the trafisfer switch is ifi *the AllSDP positioft .

Palisades Nuclear Plant B 3.3.8-3 01/20/98 05/30/99

Alternate Shutdown System B 3.3.8 BASES APPLICABILITY The Alternate Shutdown System LCO is applicable in MODES 1, 2, and 3. This is required so that the plant can be maintained in MODE 3 for an extended period of time from a location other than the control room.

This LCO is.not applicable in MODE 4, 5, or 6. In these MODES, the plant is already subcritical and in the condition of red~ced PCS energy. Under these conditions, considerable time is available to_restore necessary instrument control Functions if control room instruments or control become unavailable.

ACTIONS A Note has been included that excludes the MODE change restrictions of LCO 3.0.4. This exception allows entry into an applicable MODE while relying on the ACTIONS, even though the ACTIONS may eventually require a plant shutdown. This is acceptable due to the low probability of an event requiring this system. The Alternate Shutdown System equipment can generally be repaired during operation without significant risk of spurious trip.

Note 2 has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in

Table 3.3.8-1. The Completion Time of the inoperable channel of a Function will be tracked separately for each Function, starting from the time the Condition was entered for that Function.

A.1 and A.2 Condition A addresses the situation where the required channels of the Remote Shutdown System are inoperable. This includes any Function listed in Table 3.3.8-1 as well as the control and transfer switches.

Required Action A.1 is to provide equivalent shutdown capability within 7 days. There may be several possible means of satisfying the Alternate shutdmm capability. Required Action A.rt is to restore the channel to OPERABLE status within 69 30 days. This allows time to complete repairs on the failed channel, while maintaining alterRate monitoriRg capability in accordaRce with Action A.l .

The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.

Palisades Nuclear Plant B 3.3.8-4 01/20/98 05/30/99

l Alternate Shutdown System B 3.3.8 BASES-ACTIONS 8.1 and B.2 (continued)

If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to 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 MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.3.8.1 REQUIREMENTS

This SR applies to the startup range neutron flux monitoring channel. The CHANNEL FUNCTIONAL TEST consists of verifying proper response of the channel to the internal test signals, and verification that a detectable signal is available from the detector. After lengthy shutdown periods flux may be below the range of the channel indication. Signal verification with test equipment is acceptable.

The CHANNEL FUNCTIONAL TEST of the startup range neutron flux monitoring channel is performed once within 7 days prior to reactor startup. The Frequency is based on plant operating experience that demonstrates channel failure is rare.

SR 3.3.8.2 SR 3.3.8.2 verifies that each required Alternate Shutdown System transfer switch and control circuit performs its intended function. This verification is performed from AHSDPs C-150 and C-150A and locally, as appropriate. Operation of the equipment from the AHSDPs C-150 and C-150A is not necessary.

The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the plant can be maintained in MODE 3*

, from the auxiliary shutdown panel and the local control stations.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the

Surveillance were performed with the reactor at power .

Operating experience demonstrates that Alternate Shutdown System control channels seldom fail to pass the Surveillance when performed at a Frequency of once every 18 months.

Palisades Nuclear Plant B 3.3.8-5 01/20/98 05/30/99

BASES Alternate Shutdown System B 3.3.8 SURVEILLANCE SR 3.3.8.3 REQUIREMENTS (continued) SR 3.3.8.3 is the perfermar=1ce ef a CllANNEL FUNCTIONAL TEST every 18 mer=1ths. TRi s Survei 11 ar=1ce verifies the OPE RAB IL HY of each required cer=1trel circuit (i.e., Fur=1ctier=1s 16, 17, ar=1d 18).

This *verificatiefl is performed from the AllSDPs ar=1d locally, as apprepriate. This will er=1sure that if the cer=1trol ream becomes ir=1accessible, the plaflt cafl be mair=1tair=1ed ifl. MODE 3 frem the AllSDPs ar=1d the local cer=1trel statiMs.

The 18 meflth Frequer=1cy is based efl the Heed te perform this SurveillaHce ur=1der the cer=1ditior=1s that apply duriflg a plaflt outage ar=1d the petefltial fer afl uHplaHHed traflsieflt if the Survei 11 aHce were performed with the reacter at pewer.

Operatiflg experier=1ce has shewfl that these cemp0Her=1ts usually pass the SurveillaHce whefl performed at a FrequeHey of eHce every 18 moflths. Therefere, the FrequeHcy was cor=1cluded to be acceptable frem a reliability staHdpoiflt *

This SR is modified by a Nete which states this SR is oflly required fer Fuflctiefls 16, 17, aHd 18.

SR 3:3.8.4 A CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to the measured parameter within the necessary range and accuracy.

Perfonnance of a CHANNEL CALIBRATION every 18 months on Functions 1 through 15 ensures that the channels are operating accurately and within specified tolerances. This verification is performed from the AHSDPs and locally, as appropriate. A test of the AFW pump suction pressure alarm (Function 15) is included as part of its CHANNEL CALIBRATION. This will ensure that if the control room becomes inaccessible, the plant can be maintained in MODE 3 from the AHSDPs and local control stations .

Palisades Nuclear Plant B 3.3.8-6 01/20/98 05/30/99

Alternate Shutdown System B 3.3.8 BASES SURVEILLANCE SR 3.3.8.43 (continued)

REQUIREMENTS The 18 month Frequency is based upon the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience demonstrates that Alternate Shutdown System instrumentation channels seldom fail to pass the Surveillance when performed at a Frequency of once every 18 months. Therefore. the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by twoa Notes. Note lwh+eft states that the I SRtt is not required for Functions 16, 17, and 18; Note 2 I states that it is not necessary to calibrate neutron detectors I because of the difficulty of simulating a meaningful signal. I Wide range and source range nuclear instrument channels are not I calibrated to indicate the actual power level or the flux in I the detector location. The circuitry is adjusted so that wide I range and source range readings may be used to determine the I approximate reactor flux level for comparative purposes. I REFERENCES 1. FSAR. Section 7.4. "Other Safety Related Protection.

Control. and Display Systems"

2. 10 CFR 50, Appendix A GDC 19 and Appendix R.

1

J Palisades Nuclear Plant B-3.3.8-7 01/20/98 05/30/99

l Neutron Flux Monitoring Channels B 3.3.9 B 3.3 INSTRUMENTATION B 3.3.9 Neutron Flux Monitoring Channels BASES BACKGROUND The neutron flux monitoring channels consist of two combined source range/wide range channels, designated NI-61/63 and NI-62/64. The wide range portions, (NI-63 and NI-64) provide neutron flux power indication from < lE-7% RTP to > 100% RTP.

The source range portions, designated NI-61 and NI-62, provide source range indication over the range of 1 to 1E+5 cps.

This LCO addresses MODES 3, 4, and 5 with RO more thaR ORe

full leRgth coRtrol rod capable of beiRg withdrawR aRd the Primary CoolaRt Systems (PCS) boroR coRceRtratioR less thaR the REFUELING BORON CONCENTRATION aRd MODES 3, 4, aRd 5 with the PCS boroR CORCeRtratiOR at the REFUELING BORON CONCENTRATION.

WheR more thaR oRe full leRgth coRtrol rod is capable of beiRg withdrawR, aRd the PCS boroR eoReeRtratioR is less than the REFUELING BORON CONCENTRATION In MODES 1 and 2, the neutron flux monitoring requirements are addressed by LCO 3.3.1, 11 Reactor Protective System (RPS) Instrumentat{on. 11 I

When the plant is shutdown with RO more thaR oRe full length I*

control rod capable of being withdrawn or shutdown with the PCS boron Concentration at the REFUELING BORON CONCENTRATION, both neutron flux monitoring channels must be available to monitor neutron flux pewer. If only one section of a neutron flux monitoring channel (source range or wide range) is functioning, the neutron flux monitoring channel may be considered OPERABLE if it is capable of detecting the existing reactor neutron flux. In this application, th~ RPS channels need not be OPERABLE since the reactor trip Function is not required. By monitoring neutron flux power, loss of SOM caused by boron dilution can be detected as an increase in flux. Two channels must be OPERABLE to provide single failure ptotection and to facilitate detection of channel failure by providing CHANNEL CHECK capability .

Palisades Nuclear Plant B 3.3.9-1 01/20/98 05/30/99

Neutron Flux Monitoring Channels

BASES APPLICABLE The wide range neutron flux monitoring channels are B 3.3.9 SAFETY ANALYSES necessary to monitor core reactivity changes. They are the primary means for detecting and triggering operator actions to respond to reactivity transients initiated from conditions in which the RPS is not required to be OPERABLE. They also trigget operator actioHs to aHticipate RPS actuatioH iH the eveHt of reactivity traHsieHts startiHg from shutdowH or low power coHditioHs. The neutron flux monitoring channel *s LCO requirements support compliance with 10 CFR 50, Appendix A, GDC 13 (Ref. 1). The FSAR, Chapters 7 and 14 (Refs. 2 and 3, respectively), describes the specific neutron flux monitoring channel features that are critical to comply with the GDC.

The OPERABILITY of neutron flux monitoring channels is necessary to meet the assumptions of the safety analyses and provide for the mitigatioH of accideHt aHd traHsieHt coHditioHs detection of reduced SOM.

The neutron flux monitoring channels satisfy Criterion.~ 4 of 10 CFR 50.36(c)(2) .

LCO The LCO on the neutron flux monitoring channels ensures that adequate infonnation is available to verify core reactivity conditions while shut down.

Two neutron flux monitoring channels are required to be OPERABLE. If only one section of a neutron flux monitoring channel (source range or wide range) is functioning, the neutron flux monitoring channel may be considered OPERABLE if it is capable of detecting the existing reactor neutron flux.

For example, with the source range count rate indicator functioning properly within its range, and in reasonable agreement with the other source range, a neutron flux monitor channel may be considered OPERABLE even though its wide range indicator is not functioning.

The source range nuclear instrumentation channels, NI-1 and NI-2, provide neutron flux coverage extending an additional one to two decades below the wide range channels for use during refueling, when neutron flux may be extremely low ..

This LCO does not require OPERABILITY of the High Startup Rate Trip Function or the Zero Power Mod~ Bypass Removal Function.

Those functions are addressed in LCO 3.3.1, RPS Instrumentation.

Palisades Nuclear Plant B 3.3.9-2 01/20/98 05/30/99

Neutron Flux Monitoring Channels B 3.3~9 BASES APPLICABILITY In MODES 3, 4, and 5, with no more than one control rod capable of being fiithdrawn and the PCS boron concentration less than the REFUELING BORON CONCENTRATION or in MOciES 3, 4, and 5, with the PCS boron concentration at the REFUELING BORON CONCENTRATION, neutron flux monitoring channels must be OPERABLE to monitor core power for reactivity changes.

In MODES 1 and 2, and in MODES 3, 4, and 5 with more than one control rod capable of being withdrawn, the neutron flux monitoring channels are addressed as part of the RPS in LCO 3.3.1.

The requirements for source range neutron flux monitoring in MODE 6 are addressed in LCO 3.9.2, "Nuclear Instrumentation."

The source range nuclear instrumentation channels, NI 01 and NI 02, provide neutron flux coverage extending an additional one to two decades below the wide range channels for use .during refueling, when neutron flux may be extremely low .

ACTIONS A.1 and A.2 With one required channel inoperable, it may not be possible to perform a CHANNEL CHECK to verify that the other required channel is OPERABLE. Therefore, with one or more required channels inoperable, the neutron flux power monitoring Function cannot be reliably performed. Consequently, the Required Actions are the same for one required channel inoperable or more than one required channel inoperable. The absence of reliable neutron flux indication makes it difficult to ensure SOM is maintained. Required Action A.I, therefore, requires that all positive reactivity additions that are under operator control, such as boron dilution or PCS temperature changes, be halted immediately, preserving SOM.

SOM must be verified periodically to ensure that it is being maintained. Both re~uired channels must be restored as soon as possible. The initial Completion Time of 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 to perform SOM verification takes into consideration that Required Action A.1 eliminates many of the means by which SOM can be reduced. These Completion Times are also based on operating experience in performing the Required Actions and the fact that plant conditions will change slowly

Palisades Nuclear Plant B 3.3.9-3 01/20/98 05/30/99

Neutron Flux Monitoring Channels B 3.3.9 BASES SURVEILLANCE SR 3.3.9.1 REQUIREMENTS SR 3.3.9.1 is the perfonnance of a CHANNEL CHECK on each required channel every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . A CHANNEL CHECK is nonnally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based upon the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION. .

Agreement criteria are detennined by the plant staff and should be based on a combination of the channel instrument uncertainties including ceRtrel iselatieR, indication, and readability. If a channel is outside the criteria, it may be

an indication that the transmitter or the signal processing equipment has drifted outside its limits. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure.

Since the probability of two random failures in redundant channels in any 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period is extremely low, CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. CHANNEL CHECK supplements less fonnal, but more frequent, checks of channel OPERABILITY during nonnal operational use of displays associated with the LCO required channels.

SR 3.3.9.2 SR 3.3.9.2 is the perfonnance of a CHANNEL CALIBRATION. A CHANNEL CALIBRATION is perfonned ev~ry 18 months. The Surveillance is a complete check and readjustment of the neutron flux channel from the preamplifier input through to the remote indicators.

Palisades Nuclear Plant B 3.3.9-4 01/20/98 05/30/99

Neutron Flux Monitoring Channels B 3.3.9 BASES SURVEILLANCE SR 3.3.9.2 (continued)

REQUIREMENTS This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes.

This LCO does not require the OPERABILITY of the High Startup Rate trip function or the Zero Power Mode Bypass removal function. The OPERABILITY of those functions does not have to be verified during performance of this SR. Those functions are addressed in LCO 3.3.1, RPS Instrumentation.

This Frequency is the same as that employed for the same channels in the other applicable MODES .

REFERENCES 1.

2.

10 CFR 50, Appendix A, GDC 13 FSAR, Chapter 7

3. FSAR, Chapter 14 Palisades Nuclear Plant B 3.3.9-5 01/20/98 05/30/99

8 3.3 INSTRUMENTATION ESRV Instrumentation B 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Instrumentation B 3.3.10 BASES BACKGROUND This LCO addresses the instrumentation which provides .isolation of the ESRV System (Ref. 1). The ESRV Instrumentation high radiation signal provides automatic damper closure, using two radiation monitors. One radiation monitor is located in the ventilation system duct work associated with each of the Engineered Safeguards (ES) pump rooms. Upon detection of high radiation, the ESRV Instrumentation actuates isolation of the associated ES pump room by closing the dampers in the ventilation system inlet and discharge paths. Typically, high radiation would only be expected due to excessive leakage during the recirculation phase of operation following a Loss of Coolant Accident (LOCA). The ESRV System is addressed by LCO 3.7.13, "Engineered Safeguards Room Ventilation (ESRV)

Dampers."

Trip Setpoints and Allowable Valtles The Allowable Valtle basecl on analytical specified for the ESRV Instrtlmentation is limits. Trip setpoints in accordance ~ith the Allowable Valtle ttill enstlre that the radiological

conseqtlences of the safety analysis will be acceptable.

APPLICABLE The ESRV Instrumentation isolates the ES pump rooms in the SAFETY ANALYSES event of high radiation in the pump rooms due to leakage during the recirculation phase. The analysis for a Maximum Hypothetical Accident (MHA) described in FSAR, Section 14.22 (Ref. 2), assumes a reduction factor in the potential radioactive releases from the ES pump rooms due to plateout following automatic isolation. However, no specific value is assumed in the MHA for the actuation of the isolation. The results indicate that the potential MHA offsite doses would be less than 10 CFR 100 guidelines.

The ESRV Instrumentation satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2).

Palisades Nuclear Plant B 3.3.10-1 01/20/98 05/30/99

ESRV Instrumentation B 3.3.10 BASES LCO The LCO for the ESRV Instrumentation requires both channels to be OPERABLE to initiate ES pump room isolation when high radiation e~ceeds the trip setpoint. Operatiofi with a trip setpoifit les's cofiservati'v'e thafi the fiomifial trip setpoifit, but withifi its Allowable Value, is acceptable provided that the differefice betweefi the fiomifial trip setpoifit afid the Allowable Value is equal to or greater thafi the drift allowafice assumed for each trip ifi the trafisiefit afid accidefit afialyses.

The ESRV Instrumentation SetpointAllowable Value is specified as ~ 2.2E+5 cpm. This setpoint is high enough to avoid inadvertent actuation in the event of nonnal background radiation fluctuations during testing, but low enough to isolate the ES pump room in the event of radiation levels indicative of a LOCA and excessive leakage during recirculation of primary coolant through the ES pump room; APPLICABILITY The ESRV Instrumentation must be OPERABLE in MODES 1~ 2, 3,:

and 4. In these MODES, the potential exists .for an accident that could release fission product radioactivity into the primary coolant which could subsequently be released to the environment by leakage from the ES systems which are recirculating the coolant.

While in MODE 5 and in MODE 6, the ESRV Instruinentation need not be OPERABLE since the potential for radioactive release.~_,,,is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the 10 CFR 100 guidelines.

ACTIONS The most commofi causes of chafifiel ifioperability are failure or drift of the bistable or process module sufficiefit to exceed the Allowable Value. Ty pi call y, the drift is foufid to be small afid results ifi a delay of actuatiofi rather thafi a total loss of fuficti Ofi. This determi fiati Ofi is gefierally made durifig the performafice of a CllANNEL FUNCTIONAL TEST whefi the process ifistrumefit is set up for adjustmefit to brifig it to withifi specificatiofi. If the trip setpoifit is less cofiservative thafi the Allowable Value ifi Table 3.3.1 1, the chafifiel is declared ifioperable immediately, afid the appropriate Cofiditiofi(s) must be efitered immediately .

Palisades Nuclear Plant B 3.3.10-2 01/20/98 05/30/99

ESRV Instrumentation B 3.3.10 BASES ACTIONS The most common causes of channel inoperability are outright (continued) failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the "safe" condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards).

In the event a channel's trip setpoint is found nonconserv~tiv~

with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or RPS bistable trip unit is found inoperable, then all affected Functions provided by that channel must be declared inoperable, and the plant must enter the Condition for the particular protection Functions affected.

A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function 1 channel since each channel serves to isolate a different 1 Engineered Safeguards Room. The Completion Times of each I inoperable Function channel will be tracked separately for each I

. Function, starting from the time the Condition was entered.

Condition A addresses the failure of one or both ESRV Instrumentation high radiation monitoring channels. Operation may continue as long as action is immediately initiated to isolate the ESRV System. With the inlet and exhaust dampers closed, the ESRV Instrumentation is no longer required since the potential pathway for radioactivity to escape to the

environment has been removed.

The Completion Time for this Required Action is commensurate with the importance of maintaining the ES pump room atmosphere isolated from the outside environment when the ES pumps are circulating primary coolant .

Palisades Nuclear Plant B 3.3.10-3 01/20/98 05/30/99

ESRV Instrumentation

BASES SURVEILLANCE REQUIREMENTS SR 3.3.10.1 B 3.3.10 Perfonnance of the CHANNEL CHECK once 'every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is nonnally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

)

Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. .

Agreement criteria are detennined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel.is outside the criteria, it may be an indication that the transmitter or

the signal processing equipment has drifted outside it~ limits

The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure.

The CHANNEL CHECK supplements less fonnal, but more frequent, checks of chaDnel OPERABILITY during normal operational use of the displays associated with the LCO required channels.

SR 3.3.10.2 A CHANNEL FUNCTIONAL TEST is performed on each ESRV Instrumentation channel to ensure the entire channel will perfonn its intended function. A successful test of the 1*

required contact(s) of a channel relay may be performed by the I verification of the change of state of a single contact of the I relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL. I TEST of a relay. This is acceptable because all of the other I required contacts of the relay are verified by other Technical I Specifications and non-Technical Specifications tests at least I once per refueling interval with applicable extensions.

I Any setpoint adjustment must be consistent with 1the assumptions of the current plant ~pecific setpoint analysis analyses.

The Frequency of 31 days is based on plant operating experience with regard to-channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event.

\

Palisades Nuclear Plant B 3.3.10-4 01/20/98 05/30/99

ESRV Instrumentation B 3.3.10 BASES SURVEILLANCE SR 3.3.10.3 REQUIREMENTS '.

(continued) CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains*

operational between successive tests.. CHANNEL CALIBRATIONS must be performed consistent with the plaHt specific setpoint analysis.

The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis.

REFERENCES 1. FSAR, Section 7.4.5.2

2. FSAR, Section 14.22 Palisades Nuclear Plant B 3.3.10-5 01/20/98 05/30/99

ML18066A501

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3.3.8 APPLICABILITY

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3.7 BACKGROUND

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3.3.8 APPLICABILITY

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