ML20203F194

Proposed Tech Specs Revising TS to Be Consistent W/Planned Replacement of Current Power Range Monitoring Portion of Existing Nms

ML20203F194
Person / Time
Site:	Fermi
Issue date:	12/10/1997
From:	DETROIT EDISON CO.
To:
Shared Package
ML20203F184	List: ML20203F194 NRC-97-0105, Application for Amend to License NPF-43,revising TS to Be Consistent W/Planned Replacement of Current Power Range Monitoring Portion of Existing Nms
References
NUDOCS 9712170306
Download: ML20203F194 (40)

Text

~ s SAFETY LIMITS AND tlMITING SAFETY SYSTEM SETTINGS 2.2 L1HITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.'.1-1.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint* 1ess conservative l than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requ.irement of Specification 3.3.1 until the channel is restored to OPERABLE status with

-its setpoint adjusted consistent with the Trip Setpoint value.

9712170306 971210 PDR ADOCK 05000341 P FDR

• The APRM(4 ok41ste(Jdt.rGipefhaffneed not be declared inoperable upon enterin,1 s ngle recirculatton lobp operation provided thehsetpoints are OEi2 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per Specification 3.4.1.1. ,

810$@

FERH1 - UNIT 2 2-3~ Amendment No. 53 2

.= -

TABLE 2.2.1-1 '. ',

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE TRIP SETPOINT VALUES FUNCTIONAL UNIT

Intermediate Range Monitor, Neutron Flux - High s 120/125 divisions of s 122/125 divisions

'E l.

of full scale of fu11 scale 7 2. Average Powar Range Monitor:

Neutron Flux-Upscal s 15% of RATED s 20% of RATED

a. (Setdown) THERMAL POWER THERMAL POWER ateChermal Power-Upscale W

" b.

~' ^

27 64.3%, with F10w Blased s O M 61.4%, with s 0.63 a saximum of a maximum of '

High Flow Clamped s 113.5% of RATED s 115.5% of RATED THERMAL POWER THERMAL POWER h rjag si gle recjeculat on  ;

j dop tion:

x _

/ s 0. +5 .3%,* - 0.63W59.

/

• i

~ > Flow'Blased

/

s. / / NK b/High F,1o s lib % of RATED s 120% of RATED

c. xp Neutron Flux-Upscale THERMAL POWER THERMAL POWER

~

h-OlA- - i T Kip DOTE 5 -

,N A k 3. ReactoM1'3teali DBue1TEMig s 1093 psig - s 113 psig E a 173.4 inches

• a 171.9 inches

4. Reactor Vessel Low Water Level - Level 3 x  %

• Se Ba ure B 3/4 3-1.

?

w

• ur i 1rcu tion 100 ration er tharHidjusting thy ,A RM F1,ow'lliased Setpoints Jo P ly w) he e loo ues, th ain of the-AFRMs may bpedjuste[d fdr a period 1ot to exeed at the nal APRM d 10T rated' RATED THERl4AL POWER 4'nd a> tot w 72 houfs suc e adjustpdPRM A,ren' dings ar3 at'least 5.1

? FJtTP, prov ed tha readings do not f osted ort the reactorMtrol paneJ. / / _

$ ~of adjustment i b AVUtME 10 Weit RRerA Mimis SWtAEED T1tGIMtLPOLUDL- UfKALE MJ NMED WIAM MNT M Af A ~:

(5 TiteRifGtN CF REUltilAAD011 tt0F DUJE T-t#W (W). BW 15 DEFillED M TKOlfiEffNCE IN INDifATED D8

[ IMBJT DF DRIVE f1MJ WH1(H PR600EI llAnm EME tuW) smEn Mb LLBr Af10 st#LLE UGT

\ CtAE fuW. nW :D% TE TNO W>P TE] LAP.0W. DW B% F0L Sif1CA gr (SERARDU._

3/4.3 INSTRUMENTATION i l . . . ,

,j 3/LLI REACTOR PROTECTION SYSTEM INSTRtalENTATIQN i l LIMITING CWWITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels '

p shown in Table 3.3.1-1 shall be OPERABLE. l APPLICABILITY: As shown in Table 3.3.1-1.

i AmM: i

, s. With the number of OPERABLE channels less than required by the Minimum '

OPERABLE channels per Trip System requirement for one trip system l'

1. Within I hour, verify that each Functional Unit within the affected trip cystem contains no more than one inoperable channel or place the inoperable channel (s) and/or that trip system in the tripped condition *.

2. Ifplacingtheinoperablechannel(s)inthetrippedcondition would cause a scram, the inoperable channel (s) shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the ACTION requirod by Table 3.3.1-1 for the affected Functional Unit shall be taken.

3. If placing the inoperable channel (s) in the tripped condition

) would not cause a scram, place the inoperable channel (s) and/or

that trip system in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With the number of OPERABLE channels less than required by the Minimum i SPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped con tion within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and take the ACTION required by Table 3.3.1-1 $

- }

5 "~ h ,

~

s ny;&Bk sinm.Mm.&M.? Mate ~-Q '

• An inoperable channel need not be placed in the tripped condition where_this would cause-a scram to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the channel was first detemined to be inoperable or the ACTION required by Table 3.3.1-1 for that Functional Unit shall be taken.

• • The trip system need not be placed in the tripped condition if this would cause a scram to occur. When a trip system can be placed in the tripped condition without causing a scram to occur, place the trip system with the most ino)erable channels in the tripped condition; if both systems have the same numur of inoperable channels, place either trip system in the tripped L condition.

b FERMI - UNIT 2 3/4 3-1 Amendment No. 7),57,100

m }

)

1 c)

TNSEB T $

c. With one or more channels required by Table 3.3.1-1 inoperable in one or more APRM Fu.ictional Units 2.a. 2.b. 2.c, or 2.d: '

1, Within I hour,-verify sufficient channels remain O restore the inoperable channels to an OPERABLE

2. Within status 12 or hours,d***.

trippe ,

l

)

7NS8R.T $

• • • An inoperable channel need not In be placed these in the cases, tri)inoperable if tie ped condition where channel is this would cause a scram to occur.not restored to OPERABLE status within the requ by Table 3.3.1-1 for the functional Unit shall be taken.

i l

l

l

... 3/4.3 INSTRUMENTATION i 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION g LANCE REQUIREMENTS 4.3.1.1 Each reactor proter. lion system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

l 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of

all channels shall be prformed at least once per 18 months, EEEPT TM6lE 4.3.1.1 1, n appisestele 3 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit

• shall be demonstrated to be within its limi per 18 months. Neutron detectors are exempt from response time testing. Each at least once lt test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundani, channels in a specific reactor trip system. -

, . ~-

VJ" " *"* ' % 2 h , 2 ' '' " **"" "*"'

p gys2.s.2.6 ..e.,2.a e d 2 c-sypwrg mg :,ysreu rugrimac rests . M fawrl& 2.c . Ta*N SH" K6 pmRWED AT LEAST 6WE PEL 24 wnu. THEL.mc (1(TEM FINITilWAL WE M.

WNCn0N 2.t nWLwM.I itMVOTlN5 AtuA 111f COW 0lMNS AT E AM NE

g ng, 2. 0VT-0F.4 *TR1f Vb1ELOnMNEL 11 Unty. R1 QMonW ANNL @ M WM '

INPVT! TD M VT e

l l

L l

l l'

'The sensor response time for Reactor Vessel Steam Dome Pressure - High and Reactor Vessel Low Water Level - Level 3 need not be measured and may be assumed to be the design sensor response time.

FERMI UNIT 2 3/4 3-la Amendment No. 75, Jpp, 111 l

M F 3.3.1-1 DEACTOR PROTECTION SYSTEN INSTRLN MTATIM APPLIC WLE MINIDMI h.-. OPERATIONAL CONDITIONS-OPERABLE PER TRIP SYSTEMCHMBE1[S) a gygg .

. FUNCTIONAL. UNIT b

1. Intermediate Range Monitors (b).

Neutron Flux - High 2 3 1

a. 2 3,4(C) 5

~

3(d) 3 3 3 1 Inoperative 2

b. 3, 4 2 5 3(d) 3 3

2. Average 2.

Power Range

. utron Flux -

% @tdown Monito2 h ) 1 3(K) 1 i

i 4

b. Thermal 1 i

4 M utron Flux -

1 c.

1 Inoperative I, 2 d.

3.

(E. _ 2.-CW f 1DERE Reactor vesses steam u

_ t[71 f -- ( )

l Pressure - High 1,2(f) 2 1 ,

4. Reactor Vessel Low Water Level - 1, 2 2 ,

1

[ Level 3 , -

t

$ 5. Main Steam Line Isolation Valve - 4 4

= Closure 1(9) e 2

i tse l

_ - .. . __ _-__ __ _ . _ _ _ _ _J

TABLE 3.3.1 1 (Continued) ,

.Rf4f10R PROTECT 10N SYSTEM INSTRUMENTATION -

IMLLHOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition previded at least one OPERA 8LE channel in the same trip systes is monitoring that parameter.

(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

' (CI Unless adequate shutdown margin has been desenstrated per Specification 3.1.1, the ' shorting links' shall be removed from.ihe RPS circuitry prior to and during the time any control rod is withdrawn.*

(d) When the 'short Y ks' are removed, the Minimum OPERA 8LE Channels Per Trip System is_ _

6 IRMs and per Specification 3.9.2,13RMs.

P,f8 (f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(9) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT IN1ECRITY is nei required.

i (i) With any co foral rod withdrawn. Not applicable to control rods removed f per Specification 3.9.10.1 or 3.9.10.2.

l (N This function shall be automatically bypassed when turbine first stage j psig, equivalent to THERMAL POWER less than 305 of

. $"$E < _' ~^

! Q(k) OMJ6(6 r CMJ0ELS Sftr tFO bJ TABm 3.3.1- 1 Au N WAL AptM Thul APRM UiAWEl. h10mEC IWT T) IDTH 700 SV$tEMC , TW MWW ILEGViht) (.L e. (T ig fJ0t* cd A TeJP S/ STEM sonnetF M u.uJc6 ISIe THE S CMWELf to couetem A cw M wat m>.t. Nov>tm TEST Tim e

'"" "'*23L.1%i1Cff3&"W22.4 2*'C ** R

\

( *Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Amendment No. M,87 FERMI - UNIT 1 3/435 4/30/96 .

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

TABLE 4.3.I.1-1 z

b REACTOR PROTECTION SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRim Nis OPERATIONAL

$ CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 5 CHANNEL FUNCTIONAL UNIT CHECL TEST CALIBRATION (a) SURVEILLANCE _ REQUIRED

]

1. Intermediate Range Monitors:

a. Neutron Flux - High S/U.S,(b) S/U(C),W SA 2 S W SA 3,4,5

b. Inoperative NA W 2,3,4,5 (m)

Average Power Range Monitor ( -

2.

a. Neutron Flux - ,(b) ,)f A I

lb 2 (Setdown)

SA"I NU C EI R b. e Simula ed UPRAE F

erna ower - O g g(d) , ,p 1 O c. Neutron Flux -

F Ur5CAtE h SA

, W(d)g qm 3

d. __ NA _ _

NA 1, h (e._ laanerative

_ g *1)

~b 2-our-or-M TR P vcTERS _- ]g yg

3. Reiactor YesseTateam Dome R 1, 2 Pressure - High S Q(k) ,

i k 4. Reactor Vessel Low Water Q(k) R I, 2 E Level - Level 3 S R ~-

S 5. Main Steam Line Isolation 1 Valve - Closure R g NA Q w 6. Main Steam Line Radiation - I,2(l)

? High S Q R Q(k) R 1, 2

7. Drywell Pressure - High S

.e. . . ,

o. .- t i

TABLE 4.3.1.1-1 (Continued)

! REACTOR PROTECTION SYSTEM IllSTRINNTATION SURVEILLANCE REcularnr1ETS OPEMTIOML  !

5

• CHAlWIEL CHAfulEL FUNCTIOML CHAf81EL COWITICIIS FOR tRIICH i FUNCTIOM L UNIT CHECK- TEST CALIOMTI(NI SURVEILUWICE REGUIRED j

[! 8. Scram Discharge. Volume Mater j '

Level - High

a. Float Switch NA R 1,2,5(j)

b. Level' Transmitter S Q(k)

Q R I,2,5(j)

.9. Turbine Stop Valve - Closure' NA Q R 1

10. Turbine Control Vhlve Fast

Closure NA Q NA 1 l' 4 .

4 w

1 11. Reactor Mode Switch 1,2,3,4,5 l w Shutdown Position MA R HA

12. Manual Scram NA W NA 1,2,3,4,5 l!

( 13. Deleted.

neutron detectors any be escluded from OtunsEL CAL! ORATION. i t

i (e) i (b) The IM and SWI channels shell be determined to oorlap for et leest 5 decades dertog each starte, efter entertog SPtmilEEAL Csettlen 2 and the INI and ArnN chonnels shall be determined to everlep for et leest 5 decades eartny each controlled sketdeun, if not performed witMa the prevleue 7 deys.

_ $ Withle 24 1 nsrs prior to startup. If not performed within the previous 7 days.

f * (c)  !

o (d) This calibretten shell conslet of the adjustment of the APWI channel to confers to the power values colcoleted by a heet helence esclag 9PEIIATienet pawa , ser nr mann i--- - - - = --- ^ am. - -- - ** the eboelste difference is greeter then 21 of matts Intment  !

CenDITiOE 1 when w fim a in RIT_ $4ap3r_T=1 turn co nst k REEnfTRA8CetsflE 8

h  !

efdIts edustEent;epflie Jemme #

• _" ' MtewtET" ^M'M, i

IlbrottenGhet 'eessfR

[

? (e)

Lpans shell he ce threted et leest once per ness effective fell poner heers (irrn) estas the Tw erstem.

.g (f) - .

' . (g) ^*^2 _

6. -

g th)- @eep ^ tar W_-;sW s verMeine-M M , 1 ~= _ ~ ^ ^^ 'l Ileed numerh is .__" per1 "54 Speelfleetten 3.19.1.

. (l) This renerven is not regetree te he OPERROLE uten the reacter pressere ..

, M (j) tHth any centeel red withdrawn. Not oppilceble to centrol rode removed per Speelfleetten 3.9.10.1 or 3.9.10.2.

l!

!- (kl Inciedes__verificett_en of_the trl_s h int of the trte entt.

l .D ' CMINIEL ielOMAL IDT JIta(L 18ec(JIIE TK RAW INNT RiftuneR y RLLUgMGYW T1uuCeullets}

l (m) t7 sawn w ac paremso use E*netm was 2 swe . w i w iz nud MT R O nt A '*8E 2.

~__

i

INSTRUMENTATION I 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPEPATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consi. stent with the values shown in the Trip Setpoint column of Table 3.3.6 2.

APPLICABILITY: As shown in Table 3.3.61.

ACTION:

a. With a control rod block instrumentation channel trip setpoint* 1ess conservative than the value shown in the Allowable Values column of Table 3.3.6 2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

, b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip function requirement, take the ACTION required by Table 3.3.6 1.

SURVElltANCE PEOUIPEMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentatter channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERA 1!0NAL CONDITIONS and at the frequencies shown in Table 4.3.61.

• The APRMGWAtisecNw6cfER1Ex TED M(PQ0 FUWUl Mtilvf11sWumenfT6o

~

eed not be declared l inoperable upon entering single reactor recirc'ulat'on loop operation provided the jetpoints are @ usted1within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per Specification 3.4.1.1.

Yl!ANGED)

FERMI UNIT 2 3/4 3-41 Amendment No. J), 69 m , _ . m . _ , . _ _ . _ . _ . , _ _ , . . . . - , . _ . _ _

Ente 3.3.6-1 CONTROL ROD BLOCK INSTRt!MINTATION MINIMUM APPtICABIE

OPERfiBLE CIIAHNELS OPERATIONAL

$ TRIP TUNCTION PER TRIP Ft#iCTION CON 0!il0ftS ACTION

, I. R00 SLOCK MONITOR (a) e a. Upscale 2 1* 60 5 b. Inoperative 2 1* 60

• c. Downscale 2 1* 60

~

2. .APRM ' A W E^2# 'D" " W 9101.ATED TRANAL rower,-cfTCALE

a. (T)cE&t1U6ed'@ntf931fiLAB@) 73 1 61

b. Inoperative __ -- F 3 j

6j 3 1, 2@ 61 c.T Downscale MULATED THOl#Nf 1 61 d.Se01to6 IluY)- Upscale ( Setdown)h df,3  ? 61 3.

Q._ flDfi mGtt )

SOURCF RANGE M NTTORS

@ ( @) V@

a. Detector not full in(b) 2 61 3(I) 2 5 61 R b. Uptcale(d 2 61

• 3(I) 2 L 61 Y c. Inoperative (C) 2 61

~

3(I) 2 5 61

d. Downscale(d) 2 61 3(f) 2 5 61

4. INTERMEDIATE RANGE MONiiORS

a. Detector not full in 6 2, 5 61

b. Upscale 6 2, 5 61

c. Inoperatlye 6 2, 5 61

d. Downscalet') 6 ,

2, 5 61

5. SCRAM DISCHARGE V0ttME

a. Water level-High 2 1. 2, 5** 62

b. Scram Trip Bypass 2 2, 5** 62

6. fBEACf0R C96CXNT3YSTEM FIEfffCULATM(Fl0Wf f . 3

4. cale 2 / 62

. ara 2 62

7. REACTOR MODE SWITCH SHUTDOWN POSITION 2 3, 4 63 O

~

TABLED.L e 6-2 .

I!

l - . i CONTROL ROD BLOCK INSTRIMENTATION SETFOINTS

, t TRIP SETPOINI RLOWELE VR K j IRIP FUNCTION

- I. R00 stock MONITOR As specified in the '

. a. Upscale As specif!?* la the CORE OPERA'lleC CORE OPERATING e LIMITS IIEFORT LIMils RFART

~

b. Inoperative M M.

4Kirt RoyaRAW4M k2 94% of Reference L,evel a 92.35 of Reference Level l

c. Downscale ~

-aw

2. .

a.

W we= =/ Pbwu -vex 4Ic tw#)*

s 0.63 W + 55.6 C 7

s e.63@+ 58.M 'l I t . 3 with a maximum of Ith a maximum of ensed fFrou 084kaw5 rknac r ..o 10Eprso rm-at -)

N"# 2)! aq 53 N@ ~M [

l M M b

p ~ww rw/ h c..(-DownscaleInoperative

d. hug >- Upscalg.(tetdown) 2 5% of RATED THEIWWil FWER

~

2 35 of MTED 1MWWit. POWER s 125 of RATED THERML PONER s its of RATED TEIWWnt. FONER . --i 2 ~

c. cro~-vescale  % -

ssun,aem ne $us eg a w j 3. SOURCE ltAHEEBEINITORS m

a. Detector not full in HA f

s 1.0 x 105 cps s 1.6 x 105 cys

b. Upscale M

c: Inoperative M

e. a 3 cys** a Z cys** l' g d. Downscale .

e e W isaa - , .. ' y&PM

, QIsie

" ster be reduced to I S.T cys provided ttie sIgmel-to-metse rette g re. ,

,. c#

Flow St :pelate to JJelth 90te ."

gele of#. ,

Ng' l las rectreeletten leap on ret adjuott re dage giuIsr l emesed sto f

^

for a Fr ^

' es .

t age de IWW-ef-WlES NEWWL (my i2-- _ a^ % settee

, _ _ _ : .}s.lEp en .d._  !

g t FBfP. dad ted .l g b (conteel penet

, N$le 2kiRE AVBAE PGER.tWIGE meinA, seeUWED1NElueAL fMER.- MDKE Hsf BARD 300 blSOL SUMM 1MIET AT k IXIlk 130f DANE M8W fW1. SW li 1XF9ED E 1)E Mf1BIOKE IM luS0lEED 9tWEREN DW MRGET IF9WE FLAN BRUCR fR48DEE fuuEn . ;

CEERm) M1WEh! TWh LIIF MD JnW61E Lapr afGMaW AT DIC SaalE (8tE R81L AW 47. TR 1NDUWF SFCMast. a us s 5 9r ,

E3 E48 SEElE lAIP OMNt334 7 - - _

r i i

l l

l

la6LE t

STRUMENTAT '

! 3 CMNKL OPEMTient t CMMEL FUNCTIUML CMMNEL CO WIT 10R$ FOR W ICH

'g 1 RIP FUKIION

1. 300 BLOCK FENIITOR CMCK TEST CALIBMTIND) SURTEILLMCE KOUIKD j a. Upscale M

b. Inoperative M

,g -t z.Ye4=.s 1* i i

c. Downscale i .g m .M I*

M 4/5 3% 4 2.m I*

5A J

! 2. M a.

SIMUdtD TIKleMll P8WE1.-VPSCALE esps ____u, r/umf f W NA h M 1 Downsc " f #A

! [nEdhnRn d._IEss weg E4 4 3 2'fENLS un dt) SA

%e 2

GL_ HLH - UP5mtE N, upscale (5etdown)

D'j; h[3 i

R

3. 50L4tCERMirlRNIITMS

a. Detector not full in M

.g _

b. Upscale S/U .N M 2",5 5 S/W .N SA 2***. 5 i Y c. Inoperative M .N M

d. Downscale 5 S/W 2***. 5 #

S/U .N SA 2***. 5 '

i

4. INTElWEDIATE RAKE MIIIIImk  !

s. Detector not full in M S/W .N -M 2, 5

b. Upscale S S/W .N 34 2, 5 3 c. Inoperative M S/U .N M 2, 5

d. Dounscale 5 S/U .N SA Z, 5 i -- 5. SCRM DISCM VOLINE

! a. Water Level - High M

b. Scram Trly Bypass 4 R 1, 2, 5**

M R M 2, 5**

! E. nraidum Gmund iner arcle-SION FL Bb) j~ * '

/

g*. 7. '

, EEACTOR FIDDE SWITG SHUIDetRI F0511105 M R M' 3, 4 .

i

3/4.4 REACTOR COOLANT SYSTEM 4

... 3/4,4.1 RECIRCULATION SYSTEM R{CIRCULAT .ON l00PS L :MIT"NG CONDl' ON FOR OPERATION ,

3.4.1.1 Two reactor coolant system recirculation loops shall be in operation. .

APPLICABILITY: OPERATIONAL C0kDITIONS 1 and 2*. l EIlM

a. With one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s: r a) Place the individual recirculation pump flow controller for the operating recirculation pumi, in the Manual mode, b) Reduce T81ERMAL POWER to less than or equal to 67.2% of RATED THERML POWER, c) Limit the speed of the operating recirculation pump to less than or equal to 75% of rated pump speed, d) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit to the value for single loop operation required by Specification

2. 2.

@lWLAltDTil8WL, PWER. UPSt. ALE RAW 614WD) e) ReJc:sthe Average Power Range Monitor (APRM)AScram and Rod Block Trip Setpoints and Allowable Values ;c those applicable for single recirculation loop operatiod per Specifications 2.2.1 and 3.3.6.

f) Perform Surveillance Requirement 4.4.1.1.4 if THERMAL POWER is less than or equal to 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is less than or equal to 50% of rated loop flow.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With no reactor coolant system recirculation loop in operation while in OPERATIONAL. CONDITION 1, imediately place the Reactor Mode Switch in the-SHUTDOWN position.

c. With no reactor coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

'Se_e Soecial Test Exception 3.10.4.

FERMI UNIT 2 3/4 4 1 Amendment No. JJ,5f,EJ,57,57,109

2.2 tlHITING SAFETY SYSTEM SETTINGS ,

BASES 2.2.1 REACTOR PROTECl10N SYSTEM INSTRUMENTATION SETPOINTS ,

The Reactor Protection System instrumentation setpoints specified in Table 2.2.1 1 are the values at which the reactor trips are set for each parameter.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptatie on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

1. Intermediate Ranae Monitor. Neutron Flux Hioh The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. -The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to- accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems. '

The most significant source of reactivity changes during the power increase is due to control rod withdrawal. in crder to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 15B.4.1.2 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1%

of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed.

The results of this analysis show that the reactor is shutdown and peak power is

-limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal /gm. Based on this analysis, the IRM provides -

protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Averaae Power Ranae Monitor NWHtm RUY -UKCAR 'S OH) 9 ForoperationatlowpressureandlowflowduringSTARTUP,thek cram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power _ plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Temperature 8 coefficients are small and control rod patterns are constrained by the RWM. Of l all the possible sources of reactivity input, uniform control rod withdrawal is the rest probable cause of significant power increase.

FERMI UNIT ? B26 Amendment No. 67

l

LIMITING SAFETY SYSTEM SETTINGS 1

RASES ,  ;

i i kEACTOR PROTECTION SYSTEM INSTRUMENTATION SfTP0lNTS (Continued)

Averane. Power Range Monitor (Continued) - i Secause the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change ,

! power by a significant amount, the rate of power rise is very slow. Generally  !

the heat flux is in near equilibrium with the fission rate. In an assumed 1

uniform rod withdrawal approach to the trip level, the rate of power rise is 1 not more than 5% of RATED THERMAL POWER per minute and the APRM system would j

," be more than adequate to assure Jhutdown before the power could exceed the Safety Limit. The 15% Weutron Flux trin remains active until the mode switch is placed in the Run po:1 tion. -UP5CAli (SE1MWW)

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the

, system and therefore thg. monitors respond dj rec 11y and quickly to changes due i to transient operation,ror the case of th0MM) Neutron Flux-Upscale setpoint; i.e for a power increase, the THERMAL POWER of the fuel will be NN less than that indicated by the neutron fiuv due/'s the time constants of the

' 516 MAL e heat.tranofer associated with the fuel. (For the GMBtas43 Simtlated Thennal PoweT VE W m g g , a time constant oft 6 seconds is' W ebttE.AMAmjED !

R T E Wi m N JJd E A r efi uh@tM in order to simulate the fu@el thermal trar.sient CA

'RWSLIA1AL 'characterist cs. @Nore conservativ usnt for m m n g gretpointsas shown in Table 2.2.1-1. mum _valuesfy Tim,ULAD TH GTJL MuEL.t/P.nALE, (ANQ PIG RDU LtAnfU The APRM setpoints were selected to provloe acequate margm for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. For single recirculation loop operation, theGedamMIN f@E APRM3setpoints are based on s & W value of 8%. The A W value corrects for the difference in indicated drive flow (in percentage of drive flow which "J produces rated cortflow) between two loop and single loon operation of the

@ me core flowJhepeereastAnuitpointAsuteWwsmbraineWene tt.heMapq) t setpolft curve 4yL8%f The QHRAow111anin6tT)ow1fis#*:-mm F'

~

p d setpoint is not( D to single loop operation as core power levels g which would requir tk mit; are not achievable in a single loop CM OWD MRNht RWER-VfMAll 141G RDW &AMPED gi Cd"[Mdh"'eb9 f

[ 3. Reactor Vesse' 5 team Dome Pressure-Hiah 3 ,

High pressure in the nuclear system could cause a rupture to the nuclear p system process barrier resulting in the release of fission products. A W pressure increase while operating will also tend to increase the power of the q$ reactor by compressing voids thus adding reactivity. The trip will quiukly

' reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit nonnal operation without spurious' FERMI UNIT 2 B 2-7 Amendment No. U , # , 75

T o .

Insert ' A' The APRM System is divided into four APRM channels and four 2-out of 4 Trip Voter channels. Each APRM channel provides inputs to each of the 2 out-of-4 Trip Voter channels. The four 2-out of 4 Trip Voter channels are divided into two groups each, with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one un bypassed APRM will result in a ' half-trip

• in all four 2-out-of-4 Trip Voter channels, but no trip inputs to either RPS trip system. Therefore, any APRM Function 2.a. 2.b, 2.c. or 2.d trip from any two un bypassed APRM channels will result in a full trip in each of the four voter cchannels, which in turn results in two trip inputs into each RPS trip system.

Three of the four APRM channels and all four of the 2-out of-4 Trip Voter channels are required to be OPERABLE to ensure that no sinole failure will The 2 out-of 4 Trip Ya r includes preclude a scram on a valid signal. separate outputs The 2 out of-4 to RPS Trip Voter for the indepen function which is redundant (four total outputs).

2.e must be declared inoperable if any of its functionality applicable for the plant OPERATIONAL CONDITION is ir, operable. Due to the independent the voting of APRM trips and the redundancy of outputs, there may be conditions wherabut '

Trip VoterAPRM function 2.e isthrough inoperable, This the other functions that Trip Voter is still maintained.

may be considered when determining the condition of the other APRM functions In resulting from partial inoperability of the Trip Voter function 2.e. with the consistent addition, to provide adequate coverage of2.b, the entire and 2.ccore,least at 20 LPRM design bases for the APRM functions 2.a.from each of lhe four axial levels at inputs with at least three LPRM inputs which {he LPRMs are located, must be operable for each APidi channel.

Insert 0 The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30815P-A, ' Technical Specification Improvement Analyses for BWR Reactor Protection System," and NEDC-32410P A, ' Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option !!! Stability Trip Function," and NEDC-32410P A Supplement 1, 'NUMAC PRNM Retrofit Plus Option III Stability Trip function." The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

Insert 'B' For the digital electronic portions of the APRM Simulated Thermal Power -

U) scale and Neutron Flux - Upscale trip functions, performance characteristics t1at determine response time are checked by a combination of automatic self-test, calibration activities, and response time tests of the 2-out-of 4 Trip Voter.

l

I l

In';ert C

. . . The APRM system is divided into four APRM channels and four 2-out of-4 Trip Voter channels. Each APRM channel provides inputs to each of the four 2-out of 4 Trip Voter channels. The four 2-out-of 4 Trip Voter channels are divided into

, two groups of two each, with each group of two providing inputs to one RPS  !

tri The system is designed to allow one APRM channel, but no 2-cut- )

of p4system.

Trip Voter chtnnels, to be bypassed. Note (k) to Table 3.3.1 1 states '

tFat the Minimum Operable channels in Table 3.3.1-1 for the APRM Functional Units (except the 2 out-of 4 Trip Votei functional Unit) are the total number of APM channels required and are not on a trip system basis. The basis for I the APRM functional Unit 2.a. 2.b, 2.c, and 2.d actions is to assure trip capability within I hour and restore channel redundancy with 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4 J

W 9

- -w, - ,. -- ~-a--,- -

~e r --

o l

t J/4.3 INSTRt.HFNTATION BASES .

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUNENTATION The reactor protection system automatically initiates a reactor scram to;

a. Preserve the integrity of the fuel cladding,

b. Preserve the integrity of the reactor coolant system. ,

c. Minimize the energy which mus.t be adsorbed following a ' loss-of-coolant accident, and

d. Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform i*.s intended function even during periods when instr;;,sent channels may be out of service because of main-tenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection systen is made up of two independent ens.

There are usually four channels to monitor each parameter with two chan3els in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tri_oning of both tPJD systems will Dreduce a reactor scramJTh system s the int e IE 79 f uclea power t pret ton te s. eb es f e r REPt#E ngs he re ssed e ba for ificatic .2.1.

W INSULTc. The measurement of response time at the specified frequencies provides ssurance that the protective functions associated with each channel are

'I completed within the time lis.it assumed in the safety analyses. No credit IIM D was taken for those channels with response times indicated as applicable Response time may be demonstrated by any series of sequential overlapping or total channel test measurement, provided such tests demonst te the total channel resconse time as defined. Sensor response time v ification may be l demonstrated by either (1) inplace, onsite or of fsite st measurements, or (2) utilizing replacement sensors with certified response '

times.

ADD IMSRT"M 425hst twF Renoisemeur2 ARE .seecariep f d M W4A3L TABLE v.:. .$. {

f(kTT FK ARM 5,lmbbED MILMftL WWa.-tm'EE MD EBM b4P5 MP FW s ..

4 FERMI - UNIT 2 7 3/4 3 1 j

i.

2/4.4 REACTOR COOLANT SYSTEM BASFS w

,3, 52VtAE TiOJE POWEIL-l1PMAG Fin' 6 TEED S-3/4.4.1 RECIRCULATION SYSTEM E UD W The impact of single recire ation loop operation upon plant safety it assessed and shows that single-lo operation is permittet at power levels @ up the MCPR fuel cladding safety limit is j to 67.2% of RATED THERMAL POWER i increased asJ oted by Specificatio 2.1.2. APRM 6F#su W setpoints ([oe.DAJItinO arecadfu! rile 0as noted in Table:, 2.2.1-1 and 3.3.6-2 respectively. A time period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed to make these Mje:t.as,t.MES) following the establishment of single loop operation since the need for single loop operation often cannot be anticipated. MCPR operating limits adjustments in Specification 3.2.3 for different plant operating situations are applicable to both single and two recirculation loop operation.

To prevent potential control system oscillations from occurring in the recirculation flow control system, the operating mode of the recirculation flow control system must be restricted to the manual control mode for single-loop operation.

Additionally, surveillance on the pump speed of operating recirculation loop is imposed r.a exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below 30% THERMAL POWER or 50% rated recirculation loop flow is to prevent undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during a power or flow increase following extended operation in the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastd:sn from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is pemitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to.startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.

Amendment No. 5, 87 FERMI - UNIT 2 B 3/4 4-1

.-. Enclosure 3 to NRC-97-Ol05 PROPOSED TECllNICAL SPECIFICATION CilANGES RETYPED FORMAT Paues included 2-3 2-4 3/43-1 3/4 3-1a 3/43-2 3/435 3/43-7 3/438 3/4 3-41 3/4 3-42 3/4 3-44 3/4 3-45 3/44-1 B 2-6 B 2-7 B 2-7a B 3/4 3-1 B 3/4 3-la B 3/4 4-1

c ..

SAFETY LIMITS AND llMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING $AFETY SYSTEM SETTINGS REACTOR PROTECTION SiSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

8EPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation .,etpoint* 1ess conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

• The APRM Simulated Thermal Power - Upscale Functional Unit need not be j

- declared inoperable upon entering single recirculation loop operation provided the Flow Blased setpoints are changed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per j Specification 3.4.1.1.

1 FERMI - UNIT 2 2-3 Amendment No. EJ l

n TABLE 2.2.1-1 '

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS'

. "! ALLOWABLE

., FUNCTIONAL UNIT TRIP SETPOINT VALUES e L ' Intermediate Range Monitor, Neutron Flux - High s 120/125 divisions of s.122/125 divisions.

"' 'of full scale of full : scale -

2. Average Power Range Monitor:

[

.a. Neutron Flux - Upscale (Setdown) s 15% of RATED s 20% of RATED l '

THERMAL POWER THERMAL POWER

b. Simulated Thermal Power - Upscale l

1. Flow Biased s0.63(W-AW)*+61.4%, s0.63(W-AW)'+64.3%, . j with a maximum of with a maximum of< >

2. High Flow Clamped s 113.5% of RATED- s 115.5% of. RATED l THERMAL POWER THERMAL POWER

c. Neutron Flux - Upscale ' s 118% of RATED s 120% of RATED THERMAL POWER THERMAL POWER

'?

• 'd. Inoperative. NA NA

e. 2-out-of-4 Trip Voters NA NA -l

3. Reactor Vessel Steam Dome Pressure - high s 1093 psig s 1113 psig E

g 4. Reactor Vessel Low Water Level - Level 3 a 173.4 inches

• a 171.9 inches o.

M N *See Bases Figure B 3/4 3-1.

2 #The Average Power Range Monitor Simulated Thermal Power - Upscale Flow Biased scram setpoint varies as a '

function of recirculation loop drive flow (W). AW is defined as the difference in indicated drive flow (in-I g percent of drive flow which produces rated core flow) between two loop and single loop operation at the same l-core flow. AW = 0% for two locp operation. 4W - 8% for single loop operation. ,

E

o 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ,

LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERAB! F..

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

a. With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip System reouirement for one trip system:# l

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, verify that each functional Unit within the affected trip system contains no more than one inoperable channel or place the inoperable channel (s) and/or that trip system in the tripped condition *.

2. If placing the inoperable channel (s in the tri would cause a scram, the inoperable) channel (s) p)ed s1all be condition restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the ACTION required by Table 3.3.1-1 for the affected Functional Unit shall be taken.

3. If placing the inoperable channel would not cause a scram, place the(s) in the tripped inoperable channelcondition s and/or that trip system in the tripped condition within 12 ho(ur)s,

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Sy: tem requirement for both trip systems, place at least one trip system ** in the tripped ccadition within I hour and take the ACTION required by Table 3.3.1-1.# l

c. With one or more channels required by Table 3.3.1-1 inoperable in one or more APRM Functional Units 2.a. 2.b, 2.c, or 2.d:

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, verify sufficient channels remain OPERABLE or tripped *** to maintain trip capability in the Functional Ur.it, and

2. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, restore the inoperable channels to an OPERABLE status or tripped ***.

Otherwise, take the ACTION required by Table 3.3.1-1 for the Functional l l Unit.

1. Actions a and b not applicable to APRM Functional Units ?.a, 2.b. 2.c, and '

2.d. Action c applies only to APRM functions 2.a, 2.b, 2.c and 2.d.  !

• An inoperable channel need not be placed in the tripped condition where this would cause a scram to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the channel was first determined to be inoperable or the ACTION required by Table 3.3.1-1 fcr that Functional Unit shall be taken.

• • The trip system need not be placed in the tripped condition if this would cause a scram to occur. When a trip system can be placed in the tripped condition without causing a scram to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same ncmber of inoperable channels, place either trip system in the tripped condition.

• • • An inoperable channel need not be placed in the tripped condition where this would cause a scram to occur. In these cases, if the inoperable channel is not restored to OPERABLE status within the required time, the ACTION required by Table 3.3.1-1 for the Functional Unit shall be taken. J L FERMI - UNIT 2 3/4 3-1 Amendmentlio,JE,EJ,Jpp l

l

• • ~~

3/4.3 f TyffNTATION LIMITING CONDITION FOR-0PERATION 1 Continued)~

SUR"EILLANCE RE0UIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per-18 months, except Table 4.3.1.1-1, Items 2.a. 2.b. 2.c, 2.d and 2.e. Functions 2.a 2.b, 2.c, and 2.d do not require separate LOGIC SYSTEM FUNCTIONAL TESTS. For Function 2.e, tests shall be performed at least once per 24 months. The LOGIC SYSTEM FUNCTIONAL TEST for Function 2.e inciudes simulating APRM trip conditions at the APRM channel inputs to the 2-out-of-4 Trip Voter channel to check all combinations of two tripped inputs to the 2-out-of-4 Trip Voter logic in the Voter channels.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each applicable l reactor trip functional unit

• shall be demonstrated tc be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a s,necific reactor trip system.

• The sensor response time for Reactor Vessel Steam Dome Pressure - High and Reactor Vessel Low Water Level - Level 3 need r.ot be measured and may be assumed to be the design sensor response time.

FERMI - UNIT'2 3/4 3-la Amendment No. # , JES, JJJ,

, ag g

' TABLE"3.3.1-1" - ,

I ,

REACTOR PROTECTIOh SYSTEM IllSTIMBllATION .

,n -

jt .

APPLICA8LE-OPERATIONAL

' MINIfRM

. - FUNCTIONAL UNIT 'C00EITIONS._ OPERA PER TRIP SYSTEM 8LECHAISIEl{5)'

a ACTION

~

i-k* 1..  : Intermediatei. Range' Monitors (b). .

a. Neutron Flux - High 2 3 ;. *1 co --

g' .

3,4(C)

- 5 3(d) 3 3;

' ~

b. Inoperative 2 3  : 1-3, 4 I2i

• 5 3(d) 3 -3

. 2. : Average Power' Range Monitor, a.. Neutron' Flux ' :: Upscale (Setdown) 2 ;3(k) ,

, 1L

. b. . Simulated Thermal Power - Upscale 1 3(k) 4' '

• w- . .

1

, 3- c.. :Neetron Flux:- Upscale -I 3(k) 4-- s,

" 1:

4 ' d..  : Inoperative 1, 2 3(k)-

. e.  : 2-out-of-4 Trip Voters 1, . 2 2 1. .

3.. Reactor Vessel' Steam Dome ' '

i 1,'2(f).

Pressure - High 2 IT ,

t z4. Reactor Vessel low Water Level -- -

Level 3' 1, 2 2 -1 i 5. Main Steam Line Isolation Valve -

Closure 1(9) 4 '4

.. g x

2 a

p..

s

. a ',

1 1 -

s v .-a - e w 4 --.-! - ,r tie-+ e,- -  % e- , ~.e, .-e .+,,.-r..+..e.v e.,,,,. w .+ - m m r em -> r ,*

o l

• ^ '

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c) Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the " shorting links" shall be removed frem the RPS circuitry prior to and during the time any control rod is withdrawn.*

(d) When the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs and per Specification 3.9.2, 2 SRMs. l (e) DELETED l (f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(9) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(d) This function shall be automatically bypassed when turbine first stage pressure is s 161.9 psig, equivalent to THERMAL POWER less than 30% of RATED THERMAL POWER.

(k) Since each APRM channel provides input to both trip systems, the minimum operable channels specified in Table 3.3.1-1 are the total APRM channels required (i.e., it is not on a trip system basis). The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowed test time to complete a channel surveillance test (note (a) above) is applicable provided at least two OPERABLE channels are monitoring that parameter.

• Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

1 FERMI - UNIT 2 3/4 3-5 Amendment No. 7E, E7, i

~ '

3:- 1: *;

1. , 7 e i TABLE ~4.3.1.1-1? <

• . !;;i . ~., REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REDUIRDENTS

g

.;.  :~

s

.~ .CHANNELt ^

OPERATIONAL CHANNEL FUNCTIONAL CHAIEIEL COISITIONS FOR WICH a

{Ei FUNCT10Nat UNIT ~ CHECK- ___JIST~ CALIERATION(a): SURVEILLANCE REOUIRED

,. <1. Intermediate Range Monitors:. .

S/U(C),W' a...  ! Neutron: Flux .-: High :: S/U,S,(b)l SA 2L . s S' ' W 1SA- 3,.4, 5

' Inoperative-

^

' b.. NA W NA' 2, 3, 4,:5-

., 2.  : A,arage' Power Range' Monitor.(f):

a. -Neutron Flux -. t

[ Upscale'- (Setdown) D,(b)- :SA(m) 2jyears .2

- b. LSimulated Thermal 'i 4 Power: Upscale D SA(1) g(d),2 years (*)' I, j

c. . Neutron Flux - Upscale .D SA y(d),~2 years 17 -  !

F 4 ,

d. Inoperative NA' SA NA 1, 2 1 r

. e. 2-out-of-4 Trip Voters- D SA NA 1, 2 .

l 4 m=":i 3 .- Reactor Vessel Steam Dome I- Pressure.-LHigh. S -Q(k) R ^

1, 2

.Y

.

i 4. Reactor Vessel Low Water l 3i Level..- Leveli3' S Q(k) R z l, . 2.

W . .. r

_~ ' ..

i: i '5; Main Steam Line:: Isolation. p i- P- 1 LValve -' Closure- NA Q -R 1 l}

~

t

?- 6. ' Main Steam Line Radiationi- .  ;;

4 y

w.

High- S Q R 1,2(i) '

~

7. . Drywell Pressure' 'ligh' S Q(k).- R 1, 2 I

f.  !

,e c , ---e , , . , - ~ r , #.uw .. ., ,U., , r m:.. -.,.m e- .w.-. .. g , .~.-,,..-m _ .

r, n TABLE 4.3.1.1-1 (Continued) b REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE RECUIREMENTS C

5 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

8. Scram Discharge Volume Water Level - High

a. Float Switch NA R 1,2,5(j)

b. Level Transmitter S Q(k)

Q R 1,2,5(j)

9. Turbine Stop Valve - Closure NA Q R 1

10. Turbine Control Valve Fast Closure NA Q NA 1

11. Reactor Mode Switch Shutdown Position NA R NA 1,2,3,4,5 y 12. Manual Scram NA W NA 1,2,3,4,5 m

13. Deleted.

(a) Neutron detectors may be excluded from CHANNEL LALIBRATIDN.

3 (b) The IRM and SRM channels shall be determined to overlap for at least % decades during each startup after entering CPERATIONAL CONDITION 2 and the IRM g3 and APRM channels shall be determined to overlep for at least % decades during each controlled shutdown. If not performed within the previous 7 days.

=5 (c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> p-ior to startup, if not performed within the previous 7 days.

Q to (d) This calibration shall consist of the adjustemt of the APRM channel to conform to the power values calculated by a heat balance during 0lTRATI0aAi CONDITION 1 when THERMAL POWER 2 ?S% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERM 7.L

% POWER.

(e) Calibration includes flow input function, including flow transmitters, j

[ (f) The LPRMs shall be calibrated at least once per 1000 effectise full power hours (EFPH) using the TIP system.

(g) Deleted.

% (h) Deleted. l

? (1}

(j)

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

ba With any c' strol rod withdrawn. Not applicable to con *rol rods removed per Specification 3.9.10.1 or 3.9.10.2.

?

(k) Includes v ification of the trip setpoint of the trip unit.

(1) Channel Functional Test shall include the flow input function, excluding flow transmitters.

? (m) Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.

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

Ld ' '

INSTRUMENTATION 3/4.3.6- CONTROL' ROD BLOCK INSTRUMENTATION

. LIMITING CONDITION FOR OPERATION 3.3.6. The control' rod block instrumentation channels shown in Table 3.3.6-1 shell be OPERABLE with their trip _setpoints set consistent with the values shown'in the Trip Setpoint column of-Table 3.3.6-2.

. APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

a. With'a control rod block instrumentation channel trip setpoint*.less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to'0PERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

-b. With the number of OPERABLE channels less than required by tl.e Minimum OPERABLE Channels per Trip Function requirement, take the ,

ACTION required by Table 3.3.6-1.

f-

. SURVEILLANCE RE0VIREMENTS 4.3.6 Each of the above required control rod block trip-systems and instrumentation channels shall be demonstrated OPERABLE by the performance of

.the CHANhEL-CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Tabic.4.3.6-1.

4 S

• The APRM Simulated Thermal Power - Upscale Functional Unit need not be l

. declared. inoperable upon entering single reactor recirculation loop operation provided the Flow' Biased setpoints are changed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per l

--Spgcification 3.4.1.1.

i FERMIL- UNIT 2- 3/4 3-41 Amendment No. JJ, %),

' ~

TABLE 3.3.6-1 CONTROL R00 BLOCK INSTRUMENTATION MINIMUM APPLICABLE M OPERABLE CHANNELS OPERATIONAL E

~ PER TRIP FUNCTION CONDITIONS ACTION TRIP FUNCTION

1. R0D BLOCK MONITOR (a) 1* 60

a. Upscale 2 E 2 1* 60

% b. Inoperative 1* 60 2

m c. Downscale-l

2. AVERAGE POWER RANGE MONITOR

a. Simulated Thermal Power - Upscale 3 1 61 'j.-

3 1, 2 61 i

b. Inoperative 1 61 I l

c. Neutron Flux - Downscale 3 61 2 f,

~

l 'd. Simulated Thermal Power - Upscale (Setdown) 3 61

e. Floa - Upscale 3 1

3. SOURCE RANGE MONITORS 61 2

w a. Detector not full in(b) 3(f) 5 61 1 2 2 61

! w b. Upscale (c) i 3(f) 2 5 61

~ 2 61

c. Inoperative (c) 3(f) 2 5 61 2 61

d. Downscale(d) 3(f) 5 61

! 2

4. INTERMEDIATE RANGE MONITORS 2, 5 61

a. Detector not full in 6 l 6 2, 5 61

! b. Upscale 2, 5 61

c. Inoperatiye 6 6 2, 5 61

d. Downscalete) l S. SCRAM DISCHARGE VOLUME

a. Water Level - High 2 1, 2, 5** 62 y b. Scram Trip Bypass 2 2, 5** 62 g

o. j l

6. Deleted 3, 4 63 2

7. REACTOR MODE SWITCH SHUTDOWN POSITION

.o-TABLE 3.3.6-2 -l

,, CONTROL R00 BLOCK INSTRUMENTATION'SETPOINTS -

TRIP FUNCTION J3IP SETPOINT ALLOWABLE VALUE

, 1. ROD BLOCK MONITOR -

e a. - Upscale As specified in the As specified in the 5- CORE OPERATING CORE OPERATING-LIMITS REPORT LIMITS REPORT

[

b. Inoperative NA NA-

c. Downscale 194% of Reference Level >

._92.3% of Reference Level

2. AVERAGE POWER RANGE MONITOR- l

a. Simulated Thermal Power - Upscale

1) Flow Biased s0.63(W-AW)*+55.6%, s0.63(W-AW)"+58.5%,. '

with a maximum of with a maximum of? I '

2) High-Flow Clamped 108% of RATED THERMAL POWER 110% of RATED THERMAL POWER ~

[ b. Inoperative NA NA

c. Neutron Flux - Downscale 2 5% of RATED THERMAL POWER 2 3% of RATED THERMAL POWER '

A d. Simulated Thermal Power - k Upscale '(Setdown) s 12% of RATED THERMAL POWER s 14% of RATED THERMAL POWER I

e. Flow Upscale s 110% of rated flow s 113% of rated flow  ! .

Ee 3. SOURCE RANGE MONITORS

5. a. Detector not full in NA NA

.l' c+

b.

c.

Upscale Inoperative s 1.0 x 105 cps NA s 1.6 x 105 cps NA g: d. Downscale 2 3 cps ** 2 2. cps **

N . m.

(

• • May be reduced to 0.7 cps provided the signal-to-noise ratio 120. .. .

The Average Power Range Monitor simulated Thermal Power - Upscale Flow Blased Rod Block setpoint varies as a function of recirculation loop drive ,

N flow (W). AW is defined as the difference in indicated drive flow (in percent of drive flow which produces rated core flow) between tus loop and g

g' i

Q single loop operation at the same core flow. AW = 0% for two loop operation. AW = 8% for single loop operation.

?

O k

TABLE 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS-R E

'~

CHANNEL OPERATIONAL-CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP-FUNCTION CHECK TEST CALIBRATION (*) SURVEILLANCE REQUIRED E 1.- R00 BLOCK MONITOR a

m a. ' Upscale - NA SA 2 years-- 1*

b. Inoperative NA SA NA 1* 'l

c. Downscale NA SA 2 years 1* ,

2. AVERAGE POWER RANGE MONITOR '

I

a. Simulated Thermal Power -

Upscale - NA SA 2 years- 1

b. Inoperative NA SA .% 1, 2 l

-y

c. . Neutron Flux - Downscale NA SA 2 years 1

d. Simulated Thermal Power - I.

2 years I Upscale (Setdown) NA SA '2

e. Flow - Upscale NA SA 2 years 1 .! .

w 3. SOURCE RANGE MONITORS

a. S/U(b),W 2***, 5 h b.

Detector not full in Upscale i4A NA 2***,'5 S- S/U(b) SA

c. Inoperative NA S/U(b),W NA 2***,L5

d. Downscale S S/U(b),W,W SA 2***, 5'

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in NA S/U ,W NA 2, 5

b. Upscale S S/U ,W SA 2, 5

c. Inoperative NA S/U(b),W NA 2, 5

d. Downscale S S/U ,W SA 2,.5 N

$ 5. SCRAM DISCHARGE VOLUME

[ a. Water Level - High NA Q R 1, 2, 5**

g b. Screm Trip Bypass NA R NA 2, 5**

,E 6. Deleted j 3 7. REACTOR MODE SWITCH SHUTDOWN POSITION NA R NA' 3, 4

_** .* 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM-RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION- ,

3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*,

ACIION:

a. With one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a) Place the individual recirculation pump flow controller for the operating recirculation pump in the Manual mode, b) Reduce THERMAL POWER to less than or equal to 67.2% of RATED THERMAL POWER.

c) Limit the speed-of the operating recirculation pump to less than or equal to 75% of rated pump speed.

d) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit to the value for single loop operation required by Specification 2.1.2.

e) Change the Average Power Range Monitor (APRM) Simulated Thermal Power - Upscale flow Biased Scram and Rod Block Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specifications ? 2.1 and 3.3.6. l f) Perform Surveillance Requirement 4.4.1.1.4 if THERMAL POWER is less than or equal to 30% of RATED iHERMAL POWER or the recirculation loop flow in the operating loop is less than or equal to 50% of rated loop flow.

2._ Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With no reactor coolant system recirculation loop in operation while in OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position.

c. With no reactor coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least

. HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

• See Special Test Exception 3.10.4.

FERMI - UNIT 2 3/4 4-1 Amendment No. JJ,%f,J),$J, EJ,Jpp,

2.2 LIMITING SAFE 1Y SYSTEM SETTINGS

-BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection Systein instrumentation setpoints specified in Table 2.2.1-1 are the values at whi.:h the reactor trips are set for each parameter.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal- to or less than the drift allowance assumed for each trip in the safety analyses.

1. Intermediate Ranae Monitor. Neutron Flux - Hiah The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scaie is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal. In order to ensure that the IF,M provides the required Protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 158.4.1.2 of the FSAR.

The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak l

fuel enthalpy well below the fuel failure threshold of 170 cal /gm. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

! 2. Averaae Power Ranc.e Monitor For operation at low pressure and low flow during STARTUP, the Neutron Flux - Upscale (Setdown) scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup.

Effe;ts of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Temperature coefficients are small and control rod patterns are l constrained by the RWM. Of all the possible sources of reactivity input, l unifctm control rod withdrawal is the most probable cause of significant power increase.

FERMI - UNIT 2 B 2-6 Amendment No. # ,

.= = . -_ . ._. ..

l.'

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Averace Power Rance Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of >ower rise is very slow. Generally the heat flux is N near equilibrium with tie fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% Neutron Flux - Upscale (Setdown) trip remains active j until the mode switch is placed in the Run position.

The APRH trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation. For the case of the Neutron Flux - Upscale setpoint; j i.e., for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Simulated Thermal Power signal, a time constant of approximately 6 seconds is applied to the Neutron Flux signal in order to simulate the fuel thermal transient characteristics. More conservative Simulated Thermal Power - Upscale, Flow Biased and High Flow Clamped maximum values are used for these setpoints as shown in Table 2.2.1-1.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. For single recirculation loop operation, the APRM Simulated Thermal Power - Upscale Flow Biased setpoints are based on a AW value of 8%. The AW value corrects for the difference in indicated drive flow (in percentage of drive flow which produces rated core flow) between two loop and single loop operation of the same core flow. The Simulated Thermal Power - Upscale High Flow Clamped setpoint is not changed due to single loop operation as core power levels which would require changing this limit are not achievable in a single loop configuration.

The APRM System is divided into four APRM channels and four 2-out-of-4 Trip Voter channels. Each APRM channel provides inputs to each of the 2-out-of-4 Trip Voter channels. The four 2-out-of-4 Trip Voter channels are divided into two groups each, with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no 2-out-of-4 Trip Voter channels, to ba bypassed. A trip from any one un-bypassed APRM will result in a " half-trip" in all four 2-ott-of-4 Trip Voter channels, but no trip inputs to either RPS trip system. Therefore, any APRM Function 2.a, 2.b, 2.c, or 2.d trip from any two un bypassed APRM channels will result in a full trip in each of the four 2-out-of-4 Trip Voter channels, which in turn rqsults in two trip inputs into each RPS trip system.

FERMI - UNIT 2 B 2-7 Amendment No. EE, ES, JE, l

l LIMITING SAFETY SYSTEM SETTINGS BASES

' REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Averaae Power Ranae Monitor (Continued)-

Three of the four APRM channels and all four of the 2-out-of-4 Trip Voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. The 2-out-of-4 Trip Voter includes separate outputs to RPS for the independently voted sets of functions, each of which is redundant (four total outputs). The 2-out-of-4 Trip Voter function 2.e must be declared inoperable if any of its functionality applicable for the plant OPERATIONAL CONDITION is inoperable. Due to the independent voting of APRM trips and the redundancy of outputs, there may be conditions where the Trip Voter function 2.e is inoperable, but trip capability for one or more of the other APRM functions through that Trip Voter is still maintained. This may be considered when determining the condition of the other APRM functions resulting from partial inoperability of the Trip Voter function 2.e. In addition, to provide adequate coverage of the entire core, consistent with the design bases for the APRM functions 2.a 2.b and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel.

3. Reactor Vessel Steam Dome Pressure - Hioh High pressure in the auclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure inc. ease while operating will also tend to increase the power of the reactor by compre:; sing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal oaeration without spurious trips. The setting provides for a wide margin to tie' maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is bypassed. For a turbine trip under-these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

FERMI - UNIT 2 B 2-7a Amendment No. EJ,

-- = . .- _ =

-3/4.3 INSTRUMENTATION i e.

BASES 3/4.3.1 -REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor p_rotection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding,

b. Preserve the integrity of the reactor coolant system.

c. Minimize the energy which must be absorbcd following a loss-of-coolant accident, and

d. Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance, When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip l

, systems. There are usually four channels to monitor each parameter with two channels in each trip system.- The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system.

The tripping of both trip systems will produce a reactor scram. The APRM l sy; tem is divided-into-four APRM channels and four 2-out-of-4 Trip Voter channels. Each APRM channel provides inputs to each of the four 2-out-of 4 Trip Voter channels. The four 2-out-of-4 Trip Voter channels are divided into

two groups of two each, with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no 2-out-of-4 Trip Voter channels,. to be bypassed. Note (k) to Table 3.3.1-1 states that the Minimum Operable channels in Table 3.3.1-1 for the APRM Functional Units (except the 2-out-of-4 Trip Voter Functional Unit) are the total number

of APRM channels required and are not on a trip system basis. The basis for the APRM Functional Unit 2.a, 2.b, 2.c, and 2.d actions is to assure trip capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and restore channel rcdundancy with 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30815P-A, " Technical Saecification Improvement Analyses for BWR Reactor Protection System," and NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," and NEDC-32410P Supplement 1, "NUMAC PRNM Retrofit Plus Option III Stability Trip l- Function." The bases-for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

I f

l.

FERMI - UNIT 2 8 3/4 3-1 Amenoment No.

I 3/_L) INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION _SYSTfM INSTRUMENTATION (Continued)

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit ass imed in the safety c.nalyses. Response time requirements are specified in UFSAR T3ble 7.2-4. No credit was taken for those channels with response times indicated as not applicable except for APRM Simulated Thermal Power - Upscale and Neutron Flux - Upscale trip functions.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times. For the digital electronic portions of the APRM Simulated Thermal Power - Upscale and Neutron Flux - Upscale trip functions, performance characteristics that determine response time are checked by a combination of automatic self-test, calibration activities, and response time tests of the 2-out-of-4 Trip Voter.

1 FERMI - UNIT 2 B 3/4 3-la Amendment No. l

,. 3 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted at power levels up l to 67.2% of RATED THERMAL POWER if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2. APRM Simulated Thermal Power -

Upscale Flow Biased scram and control rod block setpoints are changed as noted in Tables 2.2.1-1 and ,.3.6-2, respectively. A time period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed to make these changes following the establishment of single loop l operation since the need for single loop operation often cannot be anticipated. MCN operating limits adjustments in Specification 3.2.3 for different plant operating situations are applicable to both single and two recirculation loop operation.

To prevent potential control system oscillations from occurring in the recirculation flow control system, the operating mode of the recirculation flow control system must be restricted to the manual control mode for single-loop operation.

Additionally, surveillance on the pump speed of operating recirculation loop is imposed to exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below 30% THERMAL POWER or 50% rated recirculation loop flow is to prevent undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during a power or flow increase following extended operation in the single recirculation loop mode.

\n inoperable jet pump is not, in itself, a sufficient reason to declare a re ulation loop inoperable, but it does, in case of a design-basis-accis *

., increase the blowdown area and reduce the capability of reflooding the et,,e; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

1 Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation i

loop following a LOCA.

l an the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal l

shock to the recirculation pump and recirculation nozzles.

FERMI - UNIT 2 B 3/4 4-1 Amendment No. EJ,E/