Text

Proposed Tech Specs Implementing Expanded Operating Domain for Plant.Expansion Allows Fuel Cyle Economics & Plant Capacity Factor to Be Enhanced Through Implementation of Flow Control Spectral Shift Operation
	ML20059H218
Person / Time
Site:	Fermi
Issue date:	08/20/1990
From:	; DETROIT EDISON CO.
To:	;
Shared Package
ML19302E206	List: ML20059H218; NRC-90-0125, Application for Amend to License NPF-43,implementing Expanded Operating Domain for Plant.Proprietary NEDC-31843P, Fermi-2 Max Extended Operating Domain Analysis, Withheld (Ref 10CFR2.790);
References
	NUDOCS 9009170047
	Download: ML20059H218 (24)
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'j . r Enclosure 3 Proposed Tec}nical Specification i.'

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LIMITING CONDITIONS FOR OPERATION AND $URVE!LLANCE REQUIREMENTS

.g . SECTION PAGE 3/4.0 APPLICABIL!TY............................................. 3/4 0-1 1/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 5HUTDOWN MARGIN........................................ 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES................................... 3/4 1 2 3/4.1.3 CONTROL R005 Control Rod Operability................................ 3/4 1 3 Control Rod Maximum Scram Insertion Times. . . . . . . . . . . . . . 3/4 1 6 Control Rod Average Scram Insertion Times. . . . . . . . . . . . . . 3/4 1 7 Four Control Rod Group Scram Insertion Times. . . . . . . . . . . 3/4 1 8 Control Rod Sc ram Accumulators. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-9 l

, Control Rod Drive Coupling............................. 3/4 1-11 Control Rod Position Indication........................ 3/4 1-13 ,

Control Rod Drive Housing Support...................... 3/4 1 15 3/4.1.4 CONTROL R00 PROGRAM CONTROLS

g. Rod Worth Min 1mizer.................................... 3/4 1-16

( Rod Sequence Control 5ystem............................ 3/4 1-17 b .

3/4 1-18 Rod Block Monitor...................................... _l 3/4.1.5 STAN08Y LIQUID CONTROL SY5 TEM.......................... 3/4 1-19 3/4.2 POWER DISTRIBUTION LIMITS 3 /4. .'.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE............. 3/4 2-1 pas.K TED 3/4 2.2 ADAN4640Hl4r. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-5 ,

3/4.2.3 MINIMUM CRITICAL POWER RAT!0........................... 3/4 2-6 ;

3/4.2.4 LINEAR HEAT GENERATION RATE............................ 3/4 2-10 L

FERMI - UNIT 2 iv i 1

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$ET10N gAfd 3/4.0 APPLICABILITY............................................ B 3/4 0-1 ,

3/4.1 REACTIVITY CONTROL SYSTEMS j F, 3/4.1.1 $ NUT 00WN MARGIN.................................. B 3/4 1-1 3/4.1.2 REACTIVITY At(0MALIES............................. B 3/4 1-1 3/4.1.3 CONTROL R0DS..................................... B 3/4 1-2 3/4.1.4 CONTROL R00 PROGRAM CONTR0LS..................... B 3/4 1-3 t-

, 3/4.1.5 STANDBY LIQUID CONTROL'$YSTEM.................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RAT E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 2 - 1 powroo 3/4.2.2 J r'. OC4MWFG. . . . . . . .,. . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 2 - 2 3/4.2.3 MINIMUM CRITICAL POWER RATIO.................... 8 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE...................... B 3/4 2 5 t

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION........ S 3/4 3-1 l 3/4.3.2 !$0LAT!0N ACTUATION INSTRUMENTATION. . . . . . . . . . . . . 8 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.................................. S 3/4 3-2 3/4.3.4 ATWS RECIRCULATION PUMP TRIP $YSTEM ACTUATION

. INSTRUMENTATION.................................. t 3/4 3-3 3/4.3.5 REACTOR CORE !$0LATION COOLING SYSTEM ACTUATION INSTRUMENTATION....................... 8 3/4 3 3 3/4.3.6 CONTROL R00 BLOCK ,

INSTRUMENTATION.........,......................... $ 3/4 3 3 3/4.3.7 MONITORING INSTRUMENTAT!0N Radiation Monitoring Instrumentation............ 8 3/4 3-3 Seismic Monitoring Instrumentation.............. 8 3/4 3-4 .

FERMI - UNIT 2 xii

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FEllMI WIT. I t-4 Amendment No. 53

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4 REACTOR PROTICTION $TITERINSTRUMINTAT10K stTfDINTS (Contin Avst)At. Power 3A.ngt,,gging (Cont.inued)

Becausethefinsdistributionassociatedwithuniformrodwithdrawalsdoesno involve high local peaks and because several rods sust be moved to chan e Mmer the rate of power rise is very slow. Senere11 tse i

by heat a significant flux is in nearamount,ilibrium equ with the fission rete. In an assumed niform rod withdrawei approach to the trip level, the rate of power rise is not een thaan 55 of RATED THERMAL p0WER per minute and the APM syster, w.op1d be more than adequate to assure shutdown before the power could exceed the $sfety Limit.

The 155 neutren flux trip reonins active until the mode switch is placed in the Run position. .

i The APRM trip syster. is calibrated using heat balance data taken during ,

steady state conditions. Fission thsteers provide the basic. input to the !

system and therefore the eenitors rescend directly and quickl 1 1.e for a ower increase the THERMAL POWER of the fuel will be less tha indicated theneutronfluxduetothetimeconstantsoftheheattransfer >oint, a associated ith the fuel. For the Flow liased Neutron Flux High set time constant of 6 a 1 seccots is introduced into the flow biand APLM in o to simulate the fuel therN1 transient chanctoristics. A sete comprvative j saximum value is used for the flow biased setpoint as shown in Table 2.2.11.

l The ApRM setgoints we selected to rovide adequate es}in for t

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&peration, W value corrects forthe reduced the difference in i between ApRM two loop andsetpoints single loop are b operation of the same core flow. The decrease in setpoint is derived by multiplying the slope of the setpoint curve by 85. The Hi h Flow Clamped Flow liased Neutron Flux.Hieh setpoint is not applicable to sin le loop operation J

as core power levels wMeh would require this limit are no achievable in a ,

single loop configuration.

3+ f.Pf!.tpy,,Vp, ssp,1,),tpaa Dome fressure-Hieh High pressure in the nuclear system could cause a rupture to tht nuclear systen, process barrier ruultino in the re1pse of fission products. A pressure increase while operating will also tend to increate the power nf the resc w by compressing voids thus adding reactivity'. The trip will quickly reduce the neutron flux counteracting the pressure increase. The trip setting is slightly higher than Ihe operating pressure to permit noral operatich without spurious 327 Amendment No. 53 FERN! - UNIT 2

2'h REACT!v!TY CONTROL SYSTEMS siOD BLO F4 MONITOR

.LIMITI4G CONDITION FOR OPERATION i

3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE. i 0^ ".!02'.'. 00!O IT;07: 1 , !,,,, Tll:T/','.' 7 L';; l , y, ,, , ...n y, c APPLICABILITY:

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a. With on9 RBM channel inoperable:

1. Verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN, and

2. Restore the inoperable RBM channel to OPERABLE status within i 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour. ,

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b. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE -

by performance of a: t

a. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies '

and for the OPERATIONAL CONDITIONS specified in Table 4.3.6-1. l

'o . CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when the reactor is operating on a LIMITING CONTROL ROD PATTERN. ,

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b c3HIRTIOt&L 00tOITION 1 with

'O. TER%L POER greater than or equal to 30% of RTED DE10RL POER and less

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than 90% of MTED THEltEL POER and the MINDG CRITICAL POER RTIO OCPR) less than 1.71, or

b. 'DEIMAL PON::R greater.than or equal- to 90% of MTED 'DEIMAL POER and the M:PR less than 1.40.

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REACTOR PROTECTION SYSTEM INSTRUENTATION SURVEILLANCE REQUIREE NTS 7

e CHANNEL OPERATIONAL CHANNEL FUNCTIONf.L CHANNEL C0fWITIONS FOR WHICH -

E FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED J

[ 8. Scram Discharge Volume Water Level - High ;

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

b. Level Transmitter S N R 1,2,5(j) !

9. Turbine Stop Valve - Closure NA M R 1 7 i- 10. Turbine Control Valve Fast j i Closure NA M MA 1 3

11. Reactor Mode Switch !

w Shutdown Position MA R NA 1, 2, 3, 4,.5 g 12. Manual Scram NA M MA 1,2,3,4,5

13. Deleted '

l T oo (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRN and SRM channels shall be detcrained to overlap for at leas. 4 decades during eact ~

startup a ker entering OPERATIONAL CONDITION 2 and the IRM and APRM cMnnels shall be determit*;d to .

oserlap for at least decades during each controlled shutdown, if not perforr.ed within the previous t 7 days. '

(c) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values i calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 2Si; of RATED i THERMAL POWER. Adjust the APRM channel if the absolute difference is greater teen 2% o' RATED THERMAL POWER fa ?JZ ch;c.c.cl jin adj;;;^ c.". x;i 8- c,M ece eith $;i ficati= . .. ._ . . ..,,. _ i

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i--1;id i- tte=i-i g th b=1 te diff;rc..,.e.

= (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a a calibrated flow signal.

l E (f) The LPRMs shall be calibrated at least once per .000 effective full power hours (EFPH) using the TIP system. ;

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i E (g) Deleted.

(h) This calibration shall consist of verifying the 6 i 1 second simulated thermal power time constant.

1 % (i) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

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

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, INSTRUMENTAT10N -

3/4.3.8 00NTROL RDB BLOCK IktTRL"L*NTATION LIMITItt CONDft!0N FDR OPERATION 1

3.3 8. The oestrol red block lastrumentatten shannels shown in Table 8.3.61 shall be OptRABLE with thefr trip setyints set eenststant with the values I

I shown in the Trip Setpoint celuun of 'able 3.366 3.

3 M t As shown in Table 3.3.5-1.

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- With a control red blest fastrumentation channel trip setpoint* 1ess eensorystive than the esise shown in the Allowable Values celuen of Table 3.3.6 2 dealere the channei inoperable until the channel is .

restoredtoOftRABLEstateswithitstripsetpointadjustedconsistent with the Trip Setpoint va' sue. -

6. With the nud er of OptRABLE channels less than required by the Minisum i

OPERABLE Channels per Trip Function requirement, take the ACTION l required h Table 3.3.61.

SRY(14L& MCI AEQUIREMENTS n

l 4.3.6 fach of the above required control red block trip star.s and

. instrumentation channels shall be demorstrated CPERABLE the performance of.

.the CHANNEL CHECK CHANNEL FUNCTIONAL ttsi and CHANNEL for the OPERATIONAL CONDITIONS and at the frequencies shown in Table ,

4.3.5-1.

b

'n L *The APAN F1sw Biesed Neutron Flux.High ::: .td O*; d s. O instrumentation need not be declared inoperabic upon enterina single reactor retirculation Specification 3.,4.1.1. loop operation provided the setpoints are adjusted FtqMI . UNIT 1 3/4 3 41 Amendmen:: No. 83 4

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BELJJJ;} (Continued) encent een steen tuttamtmAtleN

$tflalLEAftMEW15 Dettere the WM inoperable and take the ACTION pegui;wd by :

ACTION 60 -

Specifisetter.3.1.4.3. ,

trith the e mber of OptRABLE Channels: ,

ACTION $1 * * <

' a. One less then toquired by the Mintes OpttABLE thennelstesto

per Trip Function toevirone s er siste the gnoperable .

to OPERA 8Lt states within*7 l F

thannel in the tripped sensit on within the next hour.

6. Two er more less then veevived by dhe Males OptRABLE !

Channels' per Trip Function toevirement slate at least one inoperable channel in the tetyped sendIslen within 1 hovr.

the ,

With the number of OPERABLE shannels le6s then veevited 1ste j ACTI M 82.- Minimum OPERABLE Channels per Trip Function requirement. er.

the inoperable thannel in the tripped sendition within 1 With the number of OPERABLt channels less then retvited by the ACT30W 53 - Minimum DPERABLE Channels per Trip Function requirement, Smitiate a red block. .

TAttt NOTAT!n45 l wf* IMe"I j W4%-THEIMhk40 Wit ; Z M C; %;r^1 M. Rep a5*

- With more than one sentrol red withdrawn. Not applicable to sentrol rods teneved per Specification 3.9.10.1 er 3.9.10.3. ..

(4) The RBM shall be automatically typassed when a peripheral sentrol red is i selected er the referente APM cnannel indicates less than SOR of

,tAtt0 T Mt4PE L p0WIA.

(b) This function shait to automatise11y bypassed if detector'sount rete is

e 100 sps of the IM shsr. riels are en range 3 or higher.

t (c) This function shall be automatically typassed when the associated IM !

' shannels are on tange 8 or higher.

(d) This function shall be aCatically typassed when the 3M shannels are f en range 8 or higher.

(e) This function sha11 to automatically bypasses when the IM shannels are en ren,e 2.

(f) These two lovece Range Monitors shall be OPERA 8LE as required by

=-( Specification 8.9.2.

l FEMI = UNIT 2 3/4 F 43

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INSEM POR PAGE 3/4 3-43 0- When (1) DEMAL POWER is greater than or al to 30% of MTED HEIMAL POER and less than 90% of RTED HEMAL R and MCPR is less than 1.71, or (2) DEIMAL POER is greater than or e@&l to 90% of RTED DEMAL POER and E PR is less than 1.40.

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'i TABLE 4.3.6-1 (Contirued)

CONTROL. ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHAtiNEL CAllBRATION.-

(b) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior, to startup, if not performed within the previous 7 days.

J:0, 7;;:T".L . ;U:0 i 25 ef ^700 7"'T.'.' ?'"f:P. Rep (4 ce col /E /nJerP

- With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

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- - With IRHi on Range 2 or less.

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INSERP FOR PG. 3/4 3-46 o When (1) 7HE3MM, POWER is greater than or equal to 30% of RATJ!D 7tEIMAL poker and less than 90% of RTED 7tEINAL poler and MCPR is less than 1.71, or (2) DEIMM, PODER is greater than or equal to 90% of IATJiD DEJMAL P0bER and MCPR is less than 1.40.

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

ggjjgjMgt OPERATIONAL CONDITIONS 1 and t*.

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

1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 eqn1 to 705 of RATt0 THERMAL POWER.

c) Limit th6 speed of the operating racirculation pump to less than or equal to 155 of rated pump speed. ,.-

d) IncreasetheMIN!MUMCRITICALPOWERRATIO(MCpR)safetyLimitby ,

0.01 to 1.08 per Specification 2.1.2. l e) Reduce the Maximum Average planar Linear Heat Generation Rate I l (MAPLHgR) limit 0; ; = ; ;f 0.00 th;. t.t L; 7;;O;; htun -'

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.._, _, _. _. ... M:it per specification 3.2.1.

f) Reduce the Average power Range Ronitor (APRM) Scram and Rod Block Trip Setpoints and Allowable Values to

r.3";dOh;L".;..it;iainglerecirculationloopoperationiper those applicable fer ,

Specificatiotis2.t.lg4dveand3.3.6. !

, g) perform Surveillance Requirement 4.4.1.1.4 if THERMAL POWER is L less than or enval to 305 of RATED THERMAL POWER or the !

recirculation loop flow in the o equal to $05 of rated loop flow.perating loop is less than or l

2. The provisions of specification 3.0.4 are not applicable. ;

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

b. With r.o reactor coolant system recirculation loop in operation while in i 0FERAT10NAL CONDITION 1, issediately place the Reactor Modt Switch in the SHUTDOWN position. ,

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

HOT SHUTOOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Te W p*e*cTel Text [xception 3.10.4 (APRM gain adjustments sey be made in lieu of adjusting the AFRM and-48N Flow l Biased $6tpuints to comply with the single loop values for a period of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

1 FERMI - UN!Y 2 3/4 4 1 Amendment No. 53

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4.4.1.1.1 toch pump discharge valve shall be desenstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each STARTUp* prior to THEL 0EL POWER exceeding til of RATED THEREL POWER.

4.4.1.1.2 toch pump MG set scoop tube mechanical and electrical stop shall be demonstrated CPERAB,.E with overspeed setpoints 1ess than or equal .to and g."f,respectively,ofratedcoreflow,atleastonceper18 months 4.44.1.3 With one reactor coolant system recirculation loop not in operation, at least once per at hours verify thatt

a. THERMAL POWER is less than or equal to 705 of Raft 0 THERi%L POWER,and

b. The individual recirculation pump flow controller for the o >erating recirculation pump is in the Manual mode and

c. Tse speed of the operating recirculation pump is*1e,ss than or equal to 755 of rated pump speed.

4.4.1.1.4 With one reactor coolant system loop not in operatich with THERMAL POWER 1ess than or equal to 30% of RATED THERRL POWER or with recirculation loop flow in the operating loop less than or 9 qual to 505 of rated loop flow, verify the following differential temperature requiresents are set within 00 mere than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase:

a. Lars than or equal to 145'F between reactor vessel stear. space toolant and bottom head drain line coolant, and

b. Less than or equal to 50'F between the reactor cociant within the loop not in operation and the teclant in the reactor and

c. pressurirvesse1**lto50'Fbetweenthereactorcoolantwithin s

Less than or equa the loop not in operation and the operating loop.**

'If not performed within the previous 31 days. ,.'

- Requirement does not apply when the recirculation loop not in operation is isolated from the reactor pressure vessel.

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1. "Seneral Electc,e standard Application for Reactor Fuel,',NEDE-24011-P A (latestopprovedrevision). ,

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For plant operation with a single recirculation loop, the above MAPIJGR limits are multiplied by a factor specif'ed in the CDRE @ERATItC IJMITS REPORT (COLR) .

'!he COLR factor is derived from W analysis initiated from single loop operation to account for earlier boiling transition at the limiting fuel node :

/ conpared to the st;andard IOCA analysis.

I Power and flow dependent adjustn.tnts are provided in the COLR to assure that the :

fuel thermal-mechanical design criteria are preserved during almormal transients ./ t initiated from off .tated conditions. I i

X e i X x - 4 5 -

F l4 i

I a

FtRMI - UNIT 2 8 3/4 2-la Amendment No. # , $3 m

8 l

POWER DISTRIBUTION LIMITS BASES

).

' Oscarso l i 3/4,2.2 #4N-4ET4HNM vel cladding integrity Safety Limits of Specifications 2.1 were on a power ibution which would yield the design LNGR at RATE POWER.

The flow biased a d themel power-upscale scram set flow biased simulated themel power-hle control red block s of the APRM instro-ments sust bt adjusted to ensuPe%t the s not become less than the l Safety Limit MCPR or that g H pla sin does not occur in the - I degraded situation. The se ings antrted4Qock settings are adjusted in accordance with the a in this specification when,the combination of THER-MAL POWER a indicates a higher peaked power dis &tbutton to ensure that an nsient would not be increased in the degraded conditlen, a

i FERMI': UNIT-2 B 3/4 2-2 Amendment No. 42 C

maerg p/4.1.3 ._ElaimM CRIT 1t&L PO$$ RATIC (Coattaged) " -

m . m --- - 2 . - m. - - m -. --

.. x - 2 2, u : -= # 2 : r r. u : - c.n rt m i-t'g m .. . ..

m.. =_. , m...

The evaluation of a given trenstant begins with the system inittai par es shown in UF8hR Table 16.0.1 that are input to a St cora ie behavior ansient eesputer program. The sedes esed to evaluate asients are describ in 8tlTAR !!. The principal result of this eval on is the reduction in caused by the transient.

The purpose of factor of Figure 3.t.3 8 is define operating limits at other than re core f1cw tenditions. A eas then 1005 of rated flow the required MC is the product of t CPR and the Kg f actor. The.

K. factors assure that the St Limit MCPR act be violated during a flow theresse transient resulting f . setor erator speed control failure. The Kg factors any be applied to both ad automatic flow control esdes.

The K, factor values shown.in ure .3 2 were developed enetically andareapflicableto811BWR/2 WR/ and 4 reactors. The factors

. were derived using the flow trol 1 n,e corresp in to RAtt0 T RNAL POWER at rates core flow, altho tt.ty, are applicable to satended operating region.

Fot the menu flow control mode, the Kg factors were c. 91sted such that for.the maximu tow rate es. limited by the pump stoop tube se wint and the HERMAL corresgendi ativepowerwasadjusteduntiltheMCPRshanneswIthdif' POWIR along the rated flow control line tneM miting bundle s nt core flows he ratio of the McPR calculated at a given poinq of core flow, ided by e operating limit MCPR, determines the K9 .

/

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FERMI . Uutt 3 8 3/4 2 4b Amendment No. 88.42e44 40

- 1

,oeion seats

L* -

1 p0WER DISTRIBUTION LIMITS MS!$

3/4.2.3 MINIMUM CRITICAL POWER RATIO (Continued) ration in the automatic flow control mode, the same procedu em>1oyed initial power distribution was established the MCPR was eque to rating limit MCPR at RATED R and rated thermal flow.

The K, factors shown 3.2. - onservative for the Ge v s!

. Electric plant o cause the operating 1 e -f $pecifirs. m 3.2.3 er r than the original 1.20 operating limit ! % ft 'x m i

derivation fof K .* y y g ;

At THERMAL POWER-levels less than or equal to 25 percent of RATED 4&.

POWER, the reactor will be operating at minimum recirculation pump speen W %

moderator void content will be very small. For all designated control rod p w terns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial startu) testing of the plant, a MCPR evaluation will be made at 25 percent of RATED THERMAL POWER level with einimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evalua-tion below this power level will be shown to be unnecessary. The daily require-ment for calculating MCPR when THERMAL POWER is greater than or equal to 25 per-

_t- cent of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

'The requirement for calculating MCPR when a limiting control rod pattern is ap-proached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal '

limit.

3 4.2.4 LNEAR HEAT GENERATION RATE

& ., .._ ..i E s_ . ...!! $25bfE'_053

. ...... .... m v.555, $.3h .. n y.o .r;u*en;;;wwn'I5

. . . . . w .. 2$0dt i;;; ifs:th.- S m triated.

References:

1. General Electric Company Analytical Model for Loss of Coolant Analysis in Accordance with 10 CFR 50, Appendix K NEDE-20566,

-November 1975.

2. " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A, latest approved revision.

3.

Remt et brimus Ex%/e/ O >etwtly f domiN Analysu'sl' NEOG ~3/V G f ,Tuly (TTO, FERMI - UNIT 2 B 3/4 2-5 Amendment No. 42 m

e .

i Inserta Pg. B 3/4 2-5 co..

3/4.2.3 MINDEM CMTIQAL poler MTIO (Continued)

Details on how evaluations are performed, on the methods used, and how the MCPR limit is ad$isted for operation at less than rated power and flow conditions are given in Deferenoes 2 and 3 and the 00RE OPERTING IDETS REPOE.

3/4.2.4 LINEAR HEAT GE2ERTION RATE The thermal expansion rate of UOp pellets and Zircalloy cladding are different in that, during heatup, the fuel pellet could come into contet with the cla5 ding and create stress. If the stress exceeds the yield stress of the cladding material, the cladding will crmk. The IJER limit assures that at any caposure,1% plastic strain on the clad is not exceeded. This limit is a j' function of fuel type and is presented in the CORE OPERTING LIMITS REPOE.

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7 Enclosure 4 s NEDC-31843 P

' Fermi 2 Maxinum Extended Operating Domain Analysis" GE Proprietary Information

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ML20059H218

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References:

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