ML20246C946

Revised Proposed Tech Specs Re Extended Operation at Reduced Power W/Single Recirculation Loop in Operation, Reflecting Recent Industry Events Involving thermal-hydraulic Instability

ML20246C946
Person / Time
Site:	Fermi
Issue date:	08/18/1989
From:	DETROIT EDISON CO.
To:
Shared Package
ML20246C934	List: ML20246C946 NRC-89-0146, Supplemental Application for Amend to License NPF-43, Revising 880804 Application to Address Recent Concerns Re thermal-hydraulic Stability & Provide Addl Info to Proposed Tech Specs Concerning Single Recirculation Loop Operation
References
NUDOCS 8908250187
Download: ML20246C946 (42)

Text

r -

- y s2 SArtii LIMITS AIS LIMITING SAriTi SYSTEM SETTINGS

20 l

t 2.1 SAFETY LIMITS

' THERMAL POWER.-Low Pressure or Low Flow 2.1.1 ' TMERMAL POWER shall not exceed 2SE of RATED THERMAL q

' the reactor vessel steam done pressure less than 785 psig or core flow 1ess than 105 of rated flow.

APPLICABILITY: OPERATIONAL ColeITIONS 1 and 2.

,N

~ -

ACTION:

With THERMAL POWER. exceeding 25% of RATED THERMAL POWER and the reactor vessel steam done M essure less than 785 psig or core flow less than 10E cf rated flow, be in at least HDT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and -]y withAs the requirements of Specification 6.7.1.

MdY less ectrculafs'ca (vep opera h'on amf.rhe li noT b e6-THERMAL POWER. High Pressure and Hin

~ tiee Sedef U,,it MCPt.ef taog-n f

The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less with the reactor vessel steam done pressure greater than 785 psig and core 2.1.2 flow greater than 105 of rated flow.

APPLICABILITY: OPERATION _AL CONDITIONS 1 and_ M/so7 fo r tsae. rec p cy,enafief C

or leu #o tire C f*fy L*%it' MCPK of /c OG f** G I c,le she. Esfety Limit McPR c1 1c

-n-ACTION:

and the reactor vessel steam done pressure greater 4

than 785 psig and core flow greater than 10% of rated flo With MCPR less than REACTOR COOLANT SYSTEN PRESSURE The reactor coolant system pressure, as measured in the reactor 2.1.3 vessel steam done, shall not exceed 1325 psig.

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

~ APPLICABILITY:

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam done, above 1325 psig, be in at least HDT SHU the requirements of Specification 6.7.1.

8908250187 890818 PDR ADOCK 05000341

I P

PDR 2-1 FEMI - UNIT 2

ive a.-

seruii LlgtITS AIS tmaiis WETT minis maii;=*

D L19t111 M EAFtTV STtTEM SETT15 $

3.2 aratiE INTEtii6ii avevh StIRTat9ElffAT1011 SETP011tfi t

The

.eter,v.wesi

.yst intamenuti..et,.i ts enen ne =t 1

shown sa Table 2.2.1-1.

..i

.sensistent with %e Trip Setpoint v. mes As shown in Table 3.5.1-1.

ApptttAtit1TY:

=

ma.

eter pr.t.eti

.y.t int==.tets

=t,es. fin. en.t<=than the valet shown in the A1)sw g g:

declare

)y.the applicable ACT3018 statament requ

..the channel inoperable and @l the ^ahannel is vester.d ta OPERABLE status trith of $ specification 3.3.1 unti l

'-- -42o **tanint'estusted consistent with the Trip Setpoint va ue.

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FEftMI - UNIT 2 2-4 I

a-

=

3 l

Insert A

' TRIP SETPOINT ALLOWABLEVALy

1). During two recirculation

' loop operation:

a.

Flow Biased 50.58 W+59%, with 50.58W+62%,with a mariaua of a maximum of b.

High Flow Clamped

<113 5% RATED THERMAL <115.5% RATED THERMAL POWER FOWER

^

.2)' During sing 1'e recirculation loop operation:

akFlowBiased

$0.58W+54.4%,**

$0.58W+57.4%,

b.

High Flow Clamped NA NA=

During single recirculation loop operation, rather than adjusting the APRM Flow Biased Setpoints to comply with the single loop values, the gain of the APRMs may be adjusted for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that the final APRM readings are at least 5 3% of rated power greater than 1005 times FRTP, provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

I

2.1 SAFETY LIMITS S_ AMS

2.0 INTRODUCTION

_ The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel environs.

claddirig integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly

"

• Mt observable, a step-back approach is used to establish a % fatv Limit r D M ' A N b MCPR greater than 1.4Wipresents a co L 4 O Mc/A 4he MCPR i-9 ' = ! Sr '. ".

wative margin relative to the conditions required to maintain fuel claddir.g integrity. sThe fuel cladding ~is one of the physical barriers which separateThe integri the radioactive materials from the environs.

barrier is related to its relative freedom from perforations or cracking.

Although sees corrksion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally Fuel cladding perforations, however, cumulative and continuously measurable.

can result from thermal stresses which occur from reactor operation signifi-While cantly above sesign conditions and the Limiting Safety System Settings.

fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations Sipal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produceThese cond

./

\\

onset of transition boiling, MCPR of 1.0.

ficant departure from the condition intended by design for plani,4 Necation.

2.1.1 THERMAL POWER. Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated Therefore, the fuel cladding integrity Safety Limit is established by flow.

This is done by establishing a limiting condition on core THERMAL other means.

Since the pressure drop in the bypass region POWER with the following basis.

is essentially all elevation head, the core pressure drop at low power and Analyses show that with a bundle flows will always be greater than 4.5 psi.

flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 pst driving head will be greater than 28 x los 1bs/hr. Full scale ATLAS test data taken at pressures from 14.7 psis to 800 psia indicate that the fuel assembly criti-With the design peaking cal power at this flow is approximately 3.35 MWt.

factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor POWER.

pressure below 785 psig is conservative.

1 FERMI - UNIT 2 B 2-1 3

_ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _

Bases Table 82.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION t,

OF THE FUEL CLADDING SAFETY LIMIT *-

I Standard Deviation N of Point)

Guantity 1.76 Feedwater Flow

=

0.76 Feedwater Temperature 0.5 Reactor Pressure O.

Core Inlet Temperature 2.5

. Core Total Flow 3.0 Channel Flow Area 10.0

. Friction Factor Multiplier Channel Friction Factor 5.0 Multiplier 6.3 TIP Readings

{

1.5 R Factor 3.6 Critical Power uncertainty analysis used to establish the core wide Safety Limit MCPR is

• Th ed on the assumption of quadrant power symmetry for the reactor core.

i De teteof 1

'i 8 2-3 FERMI - UNIT 2

i LiWiTING SAFETY SYSTEM SETTINGS

• • M5

)

88ativil PROTECTION SYSTEM INST""ENTATION SETPOINTS (Continued -

t Averase power Ranoe Monitor (Continued)

Because the flux distribution associated with unifom red Senerally the by a significant amount, the este of power rise is very slow.In an assumed uniform heat flux is in near equilibrium with the fission rate.

rod withdrawal approach to the trfp level, the rete of power rise is not more than K of RATED THEL #EL POWER per minute and the APRM system would be more

'The 195 neutr6n flux trip remains active until ths ac6* sw

- the Run position.

'The APM trip 's stem is calibrated using heat balance data taken during Fission chambers provide the basic input to the c.tas@. state conditions.

l d quickly to changes due 2.

aystem and therefore the monitors respond direct y an to transient operation for the case of the Fixed Neutron indicated by the neutron flux due to the time constants of the heat tra casociated with the fuel. time constant of 611 seconds is introduce A more conservative to simulate the fuel themal transient characteristics.

maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.

I The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-The flow referenced trip setpoint must be adjusted by the cpecified fomula in Specification 3.2.2 in order to maintain these margins sary shutdown.

when NFLPD is greater than er equal to FATP. Adef h eW.

Reactor Vessel Steam W Pressure-Nich 3.

High pressure in the nuclear system could cause a rupture to the nuclear A pressure system process barrier resulting in the release of fission prod

~

The trip will quickly reduce the compressing voids thus adding reactivity.

The trip setting is slightly neutron flux, counteracting the pressure increase.

i t spurious

. higher than the operating pressure to permit normal operation w thou The setting ptevides for a wide margin to the asximus allowable design pressure and takes into account the location of the pressure measurement trips.

This trip to the highest pressure that occurs in the system during a For a turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

closure trip is bypassed.

k 5 2-7 FERMI - UNIT 2

p Jc p

o INSERT PAGE B 2-7 t

For single recirculation loop operation, the reduced APRM setpoints are based on a A W value of 85 The A W 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 se.me core flow. The decrease in setpoint is derived by multiplying the slope of the setpoint curve by 8%. The High Flow Clamped Flow Biased Neutron Flux-High setpoint is not applicable to single loop operation as core power levels which would require thi,s limit are no_t achievable in a single loop configuration.

h

't

__-_-_-__.__________--___-.___-_____.____________._____m._-.

.m.-.

.__m.

u L

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE t

j r

LIMITING CONDITION FOR OPERATION a ff u1 exceer}:

All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (hPLH

't:it:

3.2.1

'r:t f r-c' ^?!E"^fE *L""*" EM"^!""E :5:!' rt tr ::d 't:

d 'r! ::

-t:r ' "i;;r:: 2.2.'-1, 2.2.'-2, = f 2.2.'-3.

OPERATION &L CONDITIDN 1, when THERMAL POWER is greater than APPLICABILITY:

or equal to 25% of RATED THERMAL POWER.

ACTION:

3.2.1-1, 3.2.1-2, or 3.2.1-3, With an APLHGR exceeding the limits of Figure initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Rmib.shati he inu/h>lieel by a fac%~ of o,90}

The abcoe durirty 5 o'ng~le locp eyetofo'on.

SURVEILLANCE REQUIREMENTS 1

All APLHGRs shall be verified to be equal to or less than the limits 4.2.1 determined-from Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at b.

least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

c.

The provisions of Specification 4.0.4 are not applicable.

d.

O l

3/4 2-1 FERMI - UNIT 2 I

l

Insert 1 The IEPENGR limit which has been approved for the respective A.

fuel and lattion type as a function of the average planar esposure (as determined by the NRC g,-=.J:f asthodology described in etsTAR-II), or ~

t When hand calculations are required, the most limiting lattice 3.

type MAPIAGR limit as a function of the average planar exposure shown in the Figures 3.2.1-1, 3.2 1-2, 3.2.1-3, and-3.2.1-4 for the applicable bundle type.

4 4

l 1

- - _ _ __ =_m,m m.

m._.,.

f o.

l h

POWER DISTRIBUf!0N LIMITS 3/4.2.2 APM SETPOINTS

)

LIMITING COWITION POR OPERATION t

3.2.2 The APM flew biased neutron flux-high scram trip setpoint (5) and flow biased neutron flux-hig control rod block trip setpoint (Sg) shall be established according the following relationships:

TRIP SETPOINT ALLOWABLE VALUE E'/'b'E '##M i s (s. -- ! : "")?

1 e la ~~" ^ ~4ll H SO $ fa ~~" ^ "."")7 Eg 5, tv.i - C)!

Im.e ef 8

^

are_t '

recire 1Nreent of RATED TIENL POWER, ion flow as a percentage of t where> S and

'N=

~

flow which produces a rated core flow of 100 million 1bs/hr, at 200K of RATED THERMAL POWER T = undest value of the ratio of FRACTION OF RATED THEREL POWER divided by the EXIfRM FRACTION OF LIMITING POWER DENSITY. T is applied only if legs than or equal to 1.0 APPLICABILITY: OPERATIONAL COWITION 1, when TERMAL POWER is greater than or equal to zu of RATED THERMAL POWER.

ACTION:

With the APRM flow biased neutron flux-high scram trip setpoint and/or the d%

flow biased neutron flux-high control rod block trip setpoint less conservative U

than the value shown in the Allowable Value column for 5 or Sg, as above determined, initiate corrective action within 15 a sandadjustSand/or g o be consistent with the Trip Setpoint value thin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce l

• S t

THERMAL POWER to less then 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS U

4.2.2 The FRTP and the E LPD for each class of fuel shall be determined, the value of T calculated and the most recent actual APRM flow biased neutron y

flux-high scram and flow biased neutron flux-high control rod block trip gain readings shall be verified as indicated below,*gjusted or.the APRM setpoints verified to be within the above limits or asrequIred:

a.-

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 1 5 of RATED THE M AL POWER, and

~ Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating c.

with WLPD greater than or equal to FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

^= i w "M M ter than the FRTP during power ascension w to mar a' **Tf0~~'

THERMAL POWER; ratner -_

'j"*an the APRM maMi,- ENE~M54 gain may be adjusted such that APRM readinas== ;.

w$. --===1 to 1005 times MFLPD, rovided that

-'j.. w Mureading does not exceed i m of ^4 ") N RMAL_

(

notice of adjustment is posted on the reactor control panel.

Replace Lolth footnck 01 sert FERMI - UNIT 2 3/4 2-5 Amendment No. 9

i Insert B' l.

During tro recirculation loop operations

-t 81(0.5N+596)T S1(0.5W4624)T Sg(0.5M+50%)T Sg(0.5m+534)T -

2.

During single recirculation loop operation:

.l 81(0.5W454.#)T '

S1(0.5N+57.#)T Sg(0.5M+45.4%)T Sg(0.5M+48.8)T

~

Footnote' Insert P, age 3/4 2-5

,.~

With WLPD greater than the FR2P during power ascension up to 904 -

.of MTED 'EHEMAL PCNER, rather than adjusting the APM setpoints,

- the APM gain may be adjusted such that APM readings are greater than or equal to 1004 times WIPD provided that the adjusted APM readings do not exceed 100% of RTED THEMAL POtER and a notice of adjustment is posted on the reactor control panel. With WLPD greater than FREP and a single recirculation loop in operation, if the APM flow biased setpoints have not been adjusted to'their single loop values, then the miniman required APM reading must be increased by an additional 5.3% of rated power.

4: During single recirculation loop operation with FRIP greater than or equal to WIPD,.rather than adjusting the APM setpoints to comply with the single loop values, the APM gain may be adjusted for a period not to emoeed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that the final'AP M readings are at least 5.3% of rated power greater than 100% times FEP, provided that the adjusted APM readings do not emoeed 1004 RTED THEMAL POtER and a notice of adjustment is posted on the reactor control panel.

1 mm_______.._________.__.___.___.___m_._

_..__._,_m____...-

unu.

2/4.3.8 GANf80L man mm IIISTEReffATION G.uiis taSITION SSR SPERAfi6ii t

in Table 3.3.5-1 3.3.6. The control ved 61eck lastementation ebennels shown6h shown in the trip letpoint solen of Table 3.3.5-3.

AEP41CARILITY:Asshown.inTehle3.3.5-1.

tilth a sentre) ved histk teatrementattee abonnel trip setpoin[les MS eenservative then 18e value shown in the A11swebie values s e.

Table-3.3.6-3' declare the thennel $seperable ehtil the e withtheTrt)Setpointvalue.

With the number of OptaABLE channele less then required by the Ninfaum SPIRABLE Channels per Trip function seguirement, take the ACTION b.

required by Table 3.3.5-S.

SUWEILL:3 E RE001misiENTS d

' 4.3.6 Each of the above required control red block trip systems an instrumentation che'nnels shall he demonstrated OpfRABLE by the performan the CNANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL C#.1tR fir the OPERATIONAL COSITIONS ano et tne frequencies shown in Table 4.3

's &s MAM Flow Ramse.f Ne.stron Fluse-My emot Reof 7

neeel sof be efeeleree/

Blenk Menifer /nefreeneath,ften 1**peo*able. von eosketay singts neefer ree+s<M'on Apop apeeation povWeer fAs seipoetak **e of*(iu*9ea' within 1

h ou r.s p r foece%fte., E.Wo M,

)

3/4 3-41 FERN 1 - UNIT 2 l

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nt mg 11

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i 3]m!!

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3 nr 15 1 3 l

....;. g... $,,,lij 11 li

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' J. 1.,

n 8

~'

' ~~T ~ ~.. ~ ~

f :;

• s p

Insert C:

1 a.

Upscale

1) During two recirculation 50.66W+40%

50.66W+43%

loop operation

2) During single recirculation 50.66W+34.75

$0.66W+37 7%

loop operation Insert D:

i a.', Flow Biased Neutron Flux-High

1) During two recirculation 50.58W+50%*

$0.58W+535' loep operation

2) During single recirculation 50.58W+45.4%#'50.58W+48.4%#

loop operation a

During single recirculation loop operation, rather than adjusting the APRM and RBM Flow Biased Setpoints to comply with the single loop values, the gain of the APRMs may be adjusted for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that the final APRM readings are at least 5 3% of rated power greater than 100% times FRTP, provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

W 0

A mu e

---.._--_a...-,_---._--_

"Y

~. ~ " ' ~

W4.4 "' Tot "NT SYSTEM

- W 4.4.1 asegmaaT10N SYSTEM MG415Rff0N.100P5 Lamiini ui6W POR OPERATION 3.4.1.1 Two reacter coolant system recirculatten 1esps shall be in operstfac.

y TM' :--' :" 2 :' a ; 75, 11::. C J '. ~~ '-

1^:- 2:t ' ;.; : "! r t t::7-i...a me. -- -e eca 4. re = n a e s.t _

e 7 -; 9 OPERATIONAL 00m!T10NS b 28 APPLICABILITY:

/h.s*A E Neglets scifstWith one reacter emelant system recirculation 1eep not in oper M;

tamediately initiate action to reduce THEMAL POWER to less or and 1 to the limit specified in Figure 3.4.1.1-1 within 2 within f ate esasures to place the unit in at least MDT 1

4 12 s.

a ctor coolant systes recirculation loops i

ration, b.

With no tamediatel nitista action to reduce TERMAL to less than or ithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and equal to thir it specified in Figure 3.4.1.1-initiate mens to place the unit in at les STARTUP within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and in HDT ithin the next 6 hou With two reactor cools system recirc ion loops in operation and total core flow less the 5% of re core flow and THERMAL POWER c.

1ed i igure 3.4.1.1-1:

greater than the 11mit spec ise levels (Surveillance 4.4.1.1.3):

1.

Monitor the AP M and L o this condition and at least a)

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of ntry once per 24 he thereafte ile in this condition and, b)

Within 30 utes after the c tion of a THERMAL POWER increase at least SE of RATED RMAL POWER in an hour by to I red movement.

PM or LPRN** neutron flux noise le Is greater than ines their established baseline noise le s, tamediately With 2.

18 to within thre ate corrective action to restore the noise le in low to required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing to POWER greater than 455 of rated core flow or by redu:ing THEto less 1-1.

85ee Special Test Exception 3.10.4.

hleve19 A and C of one lpm string per core octant plus sklacaere-#"-

tored when and C of <M GL :t-daa in the center of the

--stern.

Only the center of the operating with a nonsymmetric. - --G C and two other d e :t-8aa detectors A and core LPRM strin tored for aperations with a symmetric control rod pasw....

f i

$/4 4-1 FERI - WIT 2

.... ~..

~ ~ ~ ~'

m,.i

+

Insert E i

With one reactor coolant system recirculation loop not in operation:

s.

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

j i

a) Place the individual recirculation pump flow controller fog the i

operating recirculation pump in the Manual mode.

b) Reduce THERMAL POWER to less than or equal to 70% 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 by 0.01 th.1.08 per Specification 2.1.2.

e). Reduce the Marinum Average Planar Linear Heat Generation Rate M MAPLHGR) limit to a value of 0.90 times the two recirculation loop operation limit per Specification 3 2.1.

f) Reduce the Average Power Range Monitor (APRM) Scras and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Falues to those applicable for single recirculation loop operation

• per Specifications 2.2.1, 3 2.2, and 3 3 6.

g) 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. 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With no reactor coolant system recirculation loops in operation while b.

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 seasures 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 />.

APRM gain adjustments may be made in lieu of adjusting the APRM and RBM Flow Biased Setpoints to comply with the single loop values for a period of up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

I

W.gigT_EM I

).

]

suptit t -

--m_wi s 4.4.1.1.1 geek p op elesherge selve shall be essenstrated OrthtBLE t'y 11ng each ve),e threeph et least one essplete cycle of full trevel earl STARTUP* prior to MME peter emoseflag 35 of GATED IIEWIE t

4.4.1.1.2 Eeeh pop 95 est sesop edhe eschemiset and electelset step shall be esannstrated SPER48L.E with overspeed setpetats less then er aquel to 380E and 382.N. respectively, of voted aero flow, et least once per as anoths.,

tobilsh a baselles Opm and LPWi"* eestron flem est I

(

TI M c) the voglens o is voguired unless artthin I loves of entering the

~

d in the reglen a baseltning has R4r Josseef F t

• If not perfereed within the previous 31 days.

il FEWI - WIT 2 S/4 4-2 i

l

'e----

l Insert F:

4.4.1.1.3 With one reactor coolant system recirculation loop not in, operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify thats I

a. 2HElt9J POiER is less than or equal to 70% of RTID THEINAL POiER, and

b. The individual recirculation puup flow controller for the operating recirculation punp is in the Manual mode, and

c. The speed of the operating recirculation pung is less than or equal to 75% of rate 5 pump speed.

4.4.1.1.4 With one reactor coolet system loop not in operaticn with 5HEINAL P@ER less than or equal to 30% of MTfD THEINAL PNER or with recirculation. loop tlow in the operating loop less than or

' equal to 50% of rates loop flo.t, verify the following

' deferential tauperature requirements are met within no more than 1, minutes prior to either THEINAL POiER increase or racirc'ulation flow increane:

a. Less than or equal to 145 F between remtor vessel stem space coolant and bottom head drain line coolant, and

b. less than or equal to 50 F between the rector contant within the loop not in operation and the coolant in the restor pressure vessel **, and

c. Less than or equal to 50 F between the rector coolant within the loop not in operation and the operating loop.**

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

t

b q

W COOLANT $YSTEM A L MIMP5

(

LIMITING e m nITION FOR OPERATION I

3.4,1.2 All jet p eps shall be OPERABLE.

APPLICABILITY:. OPERATIONAL CONDITIONS I and 2.

i MTigN:

Ifith one er more jet peps inoperable, he in at least HDT SHUTDOWN within g

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

,\\

SURVE1LtANCE REQUIREMENTS fMERMAL{

ANM4 4,4,},g h Ch of the above r9 quired jet pumps shall be demonstrated OPERABLE i'

Sr S '"f~'.'. ?!?

:::: ":;4L: ;' =;; i;;;;'// T

.. at least once o *a 6 per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

• by detemining recirculation loop flowhlatal core flow, and r

t kv r gid L diffuser-to-lower plenum dif rential pressure for eacgjet pop and verifying e

%thatnotwoofthefollowing Q_:g;;t;':n ;_--;; ;;;

nditions occurtwhen the gg

% 6 " - ^

;^ ^^'-

3rer r,

p g.

f O The indicated 4 circulation loop fisiw'b)

~

M Wiffers by som than 10E frcs

~

a.

'theestablishedpumpspeoploop ow characteristics.

I The indicated total core flow differs by more than 10E from the b.

established total core flow value derived from recirculation loop flow measurements.

~ The indicated diffuser-to-lower plenum differential pressure of any c.

individualjetpumpg[iffersfromthemeanofalljetpumpdifferential pressures in the semelloop by more _than 20E deviation from its noma deviation.

--;f,.

\\

nef *ff c*Ah f

a f S ec t f**cui.***

M O Li

• rc

• Tkc prov;//*o f

F/0sticled heani Het's sweeillswe 4 f er1'oe m e/ WtY/rin/2 lbou tJn fier eyeeeofioyn 25~%.of ks7ED T?fERkM

/%wfit,

%.., ??

rt test program, data shall be recorded f rs 1

e relationships.

listed to prov he criteria listed shall Comparisons of n ac he conclusion of the startup test progr I

t I

FERMI - UN]T 2 3/4 4-4

[

[-

p p

REACTOR C0OLANT SYSTEM d

RECIRCULATION PUMPS LIMITING CONDITION FOR OPERATION t

3.4.1.3" Recirculation pump speed shall be maintain *d within:

5% of each other with core flow grr,ater than or equal to 70% of a.

rated core flow, 10% of'ench other with core flow less than 70% of rated core flow.

b.

APPLICABILITY: OPERATIONAL ITIONS d 2*j ducAj two reeire=/a//an

'Joop vperc4/o'er.

With the recirculation ~ pump speeds different by more than the specified liaits, either:

Restore the recirculation pump speeds to within the specified limit a.

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or N..h".Y!*!t"..'. *. *s.t.*?.,.?. ? *. ?? *$ ?W? !* %.!*!f L..

Y r

b.

b :;:-M M. and tske the ACTION required by Specification 3.4.1.1.

y k

OMerwise., he in a f le u f Hor stw r po wd wlH,4

12. h ours.

SURVEILLANCE REQUIREMENTS Recirculation pump speed shall be verified to be within the limits 4.4.1.3 at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

"See Special Test Exception 3.10.4.

i FERMI - UNIT 2 3/4 4-5 l

f

REACTOR rYYYAMT SYSTEN-IDLE RECIRCULATION IAOP STARTUP f

LIMITING CONDITION FOR OPERATION An idle recirculation loop shall not be started unless THERMAL 3 4.1.4 POWER is within the unrestricted zone of Figure 3 4.1.4-1 and the temperature differential between the reactor pressure vessel c

steam space coolant and the bottom head drain line coolant is less than or equal to 145 F, and:

a. When both loops have been idle, unless the temperature

~

difhrehtial between the reactor coolant within.the idle loop to be started up and the coolant in the' reactor pressure vessel is less than or equal to 50 F, or c.

b. When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and 0

operating recirculation loops is less than or equal to 50 F and the operating loop flow rate is less than or equal to 50%

of rated loop flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

~With THERNAL POWER, temperature differences and/or flow rates ACTION.

exceeding the above limits, suspend startup of any idle recirculation loop.

-SURVEILLANCE REQUIREMENTS The THERMAL POWER, temperature differentials and flow rate shall 4.4.1.4 be determined to be within the limits within 15 minutes prior to startup of an idle recirculation loop.

-(

-o t

l.

g e

t y

g

_ T r

L g

W g @-

gTt 1

$i l

n e.

N W

=

g

-g

)

n'

.B s

a a

.J f

9 E

=

f E

CN 4

R g

g g

S R

2 2

(031VW %) M340d 1Whi3H13803 FEltMI - UNIT 2

~

I: '

4 p

9 -

Y i

REACTOR COM. ANT SYSTEM 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY-r t'

LIMITING CONDITION FOR OPERATION 3.4.10 The Reactor core shall not be operated in Region A'or Region-B of Figure 3.4.10-1.

APPLICABILITY:-

OPERATIONAL CONDITION 1 With the Reactor operating in Region A of Figure ACTION:

a.

3.4.10-1 immediately place the Reac+or Mode Switch in the SHUTDOWN position.

b.

Witd the Reactor operating in Region B of Figure 3.4.10-1. immediately initiate action to exit Region B by inserting control rods, If, while exiting Region B. core thermal hydraulic c.

instability occurs as evidenced by APRM readings oscillating by greater than or equal to 10% of RATED THERMAL POWER. peak to peak or LPRM rgadings oscillating greater than or equal to 30 watts /cm peak to peak, immediately place the Roactor Mode Switch in the SHUTDOWN position.

SURVEILLANCE REQUIREMENTS 4.4.10.1 The provisions of Specification 4.0.4 are not applicable.

4.4.10.2 With THERMAL POWER greater than 30% of RATED THERMAL POWER and Core Flow less than 50% of Rated Core Flow.

verify that the Reactor core is not operating in Region A or Region B of Figure 3.4.10-1 at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1 f

4 4

e>

I h

(

R t

M r.

4 e

9 em E

g

,3 k

3 8 7 c

g y

Q 4

amou enes ssp

~

W

,4

=

w g

m M 8883 %$F c

g o

e C

\\

1 4g

~

a g

s w

~

e4 5

4 o

• g i

e Tm g

\\i R

8 s

e

~

g g

e h w w unnodwmuan FERMI - UNIT 2 374 4 5

p

p j

..e s

.~

i SPECIAL TEST EECEPTIONS' j

3/4.10.4 RECIRCULATION LOOF 8 LIMITIRC CORDITION FOR OPERATION t

i i

3.10 A 'The requirements of Specification 3.4.1.1 and 3.4.1.3 that

~

recirculation loops be in operatfor with matched pump speed may be:

suspended for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the performance of PEYSICS TESTS, provided that THERMAL POWER does not exceede 31 ef RATED TdERMAL POWER.

l 1

APPLICA5fL11T: OPERATIONAL CONDITION 2. during PETSICS TESTS

~ ACTION:

~

~.

as With'the above specified time limit exceeded, insert all control rods.

b.

With the above specified THERMAL FOWER limit exceeded during PHYSICS TESTS. immediately place the reactor sofe switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at least once per hour during PHYSICS TESTS.

4.10.4.2 THERMAL F0WER shall be determined to be less than $1 of MATED THERMAL POWER at least once per hour during FHYSICS TESTS.

1

_j

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL RODS The specifications of this section ensure that (1) the minimum SHUTDOWW t

MARGIN is maintained, (2) the control rod insertion times are consistent with "those used in the safety analyses, and (3) limit the potential effects of The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive crit the rod drop accident.

The re'quirements operation.

on total rod worth and scram shape will be kept to a minimum.for problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasohable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully inserted position are consistent with the SHUTOOWN MARGIN requirements.

The number of c,nimi rods permitted to be inoperable could be more than the eight allowed oy the sp*cification, but the occurrer)ce of eight inoperable uit be Attug dwn rods could be ind!cative of a ger.sric problem and the reactor g fgg g,%,y gcg for investigation and resolution of the problem.

~ '

The control red system is designed t: bring the reactor subcritical at a during the rate fast enough to prevent the MCPR from becoming less tha

. This analysis limiting power transicni s alyzed in Section 158 of the FSARscram with the shows that the negative reaciivity rr.tes resulting from specifications, provide the average response of all the crives r.s given in the The occurrence of required protection and MCPR remairs greater than scram times longer then those spc:ified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance inte is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be

)

available when needed to accept discharge water from l

when required.

Control rods with inoperable accumulators are declared inopera Specification 3.1.3.1 then applies. accumulators that would result in less has been analyzed even thougn control rods with inoperable accumulators may Operability of the still be inserted with normal drive water pressure.

accumulator ensures that there is a means available to insert even under the most unfavorable depressurization of the reactor.

N J

0 3/4 1-2 FERMI - UNIT 2

.. 7

.g.

WL2 SAERM p LEKf8 The specifiesttens of efs sectica escure that h peak stedflag tagerature petiewlas tem moetetesad eneses tests lese-et soetest aestdant g

will est emneed the 2000T ttott spectflee to 38 tra 80.46.

ML2.1 mpaAE PUNIhR LIIEAR WAT m?Im AfLTE The peak tieddfag temperature (PCT) Sp11ewleg a poeteleted less-of-coolan eecident is primertly a fonetten of ete esseegs host generet The past clad en tan ved te ved pesar afsteshetten withis en ensembly.a tsWR for tem highest pow temperature is solem1sted essant corrected for esastfleetten. This USR accus) ta er less teen tem Asefentiens 1.tr is used ta tair seste ende along with The Technical state gap eenductance and rette-red local peeklas factor.Spectffc The ved civised ty its local peektes faster.2.2.1-1, 3.2.1-2 and 3.2.

LNet of the highest p

itettias eslue for ie atmun la Figures 6-t,

.2. - p 3.7.18E14*basid EnT14d589f%eilant' culatt s T2.

The snelysis has perforudd estag General Electric (

M r#R M.

I ans1 sensistent wfth the requirements of A semplete fan of each eede se #1ayed la the ans 4s presented in s

tous enslyses can forence 1.

s ta this ans is esapa broenn doun'as 4,,.,,

.' lead _.Chenets ffictants ta tfa saperfsati 1.

Corrected Wapert feels the NFL900 serfected.

,eerrelatten ss erese la the eere accerete typass areas = The ttue.

2. ' 1 de more vocaloutotes asfas a more eseurete

'errected guide tehe teorest resistence.

C

--m.,_--+

2

• .,m g...w

sw n

hts N

8 3/4 2-1 FElst! - est!T 2 l

ppER91HRIBEIRK1IIH75

~

afraAar MANAR Linean saAT mRegeAT10x RATE (Continued) n t

1.

Core CCFL pressure differential - 1 pel - Incorporate empth n that flew from the typ6ss to louer plenum must ove 1 psi lI ressure drop in core.

t 2.

I rate IRC pressure transfer assumpti he assumption used in E-REFL900 pressure transfer pressure is incross lag was tha

- s A few of the change t the t calculatten irrespective of CCFL. These changes are li 1. r

,g l

a.

Iriad_Charine

- j.."

1..

Sreak A' rises -

break area calculated more occurately.

~

l h.

nodei cha

/

1.

rowed Radiation and Conduction Calculation corporation of

\\

t CMASTE 05 for heatup calculation.

i lan AJ gtAfD ' =<rliisaaselen

_ i.T.

irr s ys presentes in seses iso'le' 8h 3/4.2.2 APRM SETPOINTS

~

The fuel cladding integrity Safety Limits of Specification 2.1 were based on s' power distribution which would yield the design IMBR at RATED THERMAL The flow biased simulated thermal power-agscale scram setting and flow POWER.

-biased staulated thermal power-upscale control red block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than g

The l

1,66 or that g 12 plastic strain does not occur in the degrand situation.

(M. M.

cram settings and red block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL Pt*1ER and MFLPD indicates a higher peaked power distribution to ensure that an LMGR transient would not be increased in the degraded coodition.

t

)

FERMI - UNIT 2 8 3/4 2-2

6 7

MIES TABLE B 3.2.1-1

$I9tIFICANT DPW PARNETERS TD flE t

pl$5-0F-C00UWTACCIDENTAMLYSIS Plant Parameters:

M30 ftft* dich

.. J::-i to Core flEMAL P0ER.................... 1955 of rated steam flow Wessel Staas l'atput................... 14.06 x 108 tha/hr dich cor ::conds to 105% of reted

,staam 'f sw Vesset Steam Dome Pressura............

'1055 psia Design Basis circulation Line

,r' treak Area for:

,Y a.

Large 8twaks 4.1 ft8 6.

Small treaks 0.1 ft Fuel Parameters:

PEAK TECHNICAL INITIAL

$ SPECIFICATION DESION MINIMUN LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE SENERATION RATE PEAKING POWER FUEL TYPE GE0 METRY (kW/ft)

FACTOR RATIO Initial core sx5 13.4 1.4 1.15

• r "MM N 6s5 -

's'EY ^~

~f(

ig'r*

A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 6.3 of the M.uf544.

• This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE Italt.

l i<ech e u ta rt,n to 'ep app e ro, tie,,, r.s.,

o f,, u e tea 1.

n Fe, ring e bo r'le'n3 is a.n u m e/ a f' O I secod a ffre 1-Oc d reg awffe s s of initia(

McPQ e

8 FERNI - UNIT 2 8 3/4 2-3

j POWER DISTRIBUTION LIMITS

-]

BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of-4,46, and an ane';-is of abnomal operationti transients. For any abnormal operating transients analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the =esulting MCPR does not decrease below the Safety Limit MCPP at any time during the transient assuming instrument trip setting given' in Specification 2.2.

1 To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reductio'n in CRITICAL POWER RA110 (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. ?M liettina tre rient yie!ds

+A= 1a*gast d:lt: "C"".

S:tr :dd:d t: th 5:f ty Lf it "c h Of 1.a', th ::; f r:d ei ' r

tin; 'imit "CP" Of 5;::!'keti:n 3.2.3 h eteired trd pr~e-+ed da Ff;;r: 3.2.2-1.

Aeld Jwe t 4 --T M. Trhis IS".o-O.

The evaluation of a transieni uwvini with the system initial parametersshowninfid"give.dh 15".^-1 that are input to a GE-core dynamic behavior transient computer program. TM ::d: :::d40-ev:h:t: pr ::uri::tien b ude 020 t0 h d:::Tihd 'T "E"^-211

} Ord th PT;s70: 200d '^ :: ?r;;;;7II th V

rents is de: rih d 8- "E^^-In g 2), yg ;;gp;g; ;f g3g ;7;;7= ;3;; ;$;g db wlute.

)the 8-iti:1 MC"" f: = th: input f r fertb r :::1y::: ef t h th r:1?y 'feiti g ce,vlexfj evndle with t h ti g h channel tren!! eat the-ee1 hyde ="H e m e rada daterihad h ]de/

are /rece i

utgrq5 = f4)

The principal result of this evaluation is the reduction in

/ MCPR caused by the transient.

The purpose of the K factor of Figure 3.2.3-2 is to define operating f

1 limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and the K factor. The K f

f factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor generator speed control failure. The K factors may be applied to both manual and automatic flow controi modes.

g The K, factor values shown in Figure 3.2.3-2 were developed generically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The K, factors were derived using the flow control Ifne corresponding to RATED THERMAL POWER at rated core flow, a tthenk #ey are alJo a g ir. Ale for tse exteded yrrd% reyh -

For the manual flow control mode, the K factors were calculated such that f

for the maximum flow rate, as limited by the pump scoop tube setpoint and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, detemines the K.

},

f FERMI - UNIT 2 8 3/4 2-4

Insert 3 The Te&nical specification Burimst value is the most limiting oosposite of the fuel mechanical design analysis MAPESER and thet BCCS MhP1aGR.

Mechanical Design Analysis:

NRC approved methods

' Fuel (specified in Reference 1) are used to demonstrate that all fusi rods in a lattice, operating at the bounding power meet the fuel design limits specified in Reference history, bounding power history is used as the basis for the

~

1.

Wats fuel design analysis EEPIAGR value.

Iach Analysis:

A,IDcA analysis is performed in accordance with 10CFR50 Appendix K to demonstrate that the MAPIAGR values The comply with the ECCS limits specified in 10CFR50.46.

analysis is performed for the most limiting break sise, break location, rand single failure combination for the plant.

only the most limiting MAPINGR values are shown in the Technical figures for multiple lattice fuel.

When hand specification calculations are required, these Technical specification MAPIMGR figure values for that fuel type are used for all lattions in that bundle.

For some fuel bendle designs MhPIR3R depends only on bundle type Other fuel bundles have MAPIBGRs that vary axially and burnup.

depending upon the specific combination of enriched uranius and gadolinia that comprises a fuel bundle cross section at a particular axial mode.

Each particular combination of enriched uranium and gadolinia, for these fuel bundle types, is called a lattice type.

These particular fuel bundle types have MAPIRGRs that vary by lattice (axially) as well as with fuel burnup.

= w

(

Rafarance

" General Electric Standard Application for Reactor Fuel,"

1.

NEDE-24011-P-A (latest approved revision).

F, p la.d ope ra f>*o n wtsa sin 9 e. ope ra %5 re d-c~la % r f

m e k,p keef A y 0 90 -

j lo opi & a b en MPul&4 tem its we a na lyds l i

cess fut fac1* r

• f 0 fo is c4r/M Geou LocA lThe

.s.n9 c loop sp era %

fo a cc ou df t'o r eerfier

/

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Ensett van inoperable jet pump is sut, fa itself, a sufficient reason to declare is-accleent, o recirculatten leap insperable, but it eses,in case of a east as the cere; tacrease the blandown arms and teente the capability of ref) thus, the requirement for shutepwn of the facility with a jet pump inspetable.

. Jet peep faiTure can he detected by eenitoring jet pump perfemance on a i

prescribed sch6dule for afsnificant degradation.

Recirculation pipe speed missesch limits are in essy11ance with the ECCS LOCA analysis design critarisw fee ywe ree/cc.

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14.p speermWea. Jaserf //

In order to prevent undue stress en the vessel assales and bottom head region, the recirculation leap temperatures shall he within 90'F of each other The loop temperature must aise to within prior to startup of an idle 1 esp.

. $0*F of the reacter pressure vessel coolant temperatur

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__._._.m___,

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w-

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- Jaserf =2:-

w vwg 3/4.4.2 ShFETY/ RELIEF VALVES the reactor coolant system from being pressurized abeue the A total of 11 OPERA 8LE safety /

1125 psig in accordance with the ASME Code. relief valves is required to Itait olloweele values for the worst case apset transient.

Benenstration of the safety / relief valve 11ft settings will occur only durin0 shuteswn and will be performed in accordance with the previsions of Specification 4.0.5.

The low-low set system ensures that a potentially high thrust lead (desig-noted as lead. case 0.3.3) en the ser discharge lines is eliminated du point of two valves and lowering the apening setpoint of two valves following the sequent actuations.

Sufficient roeundancy is provided for the low low set system such that failure of aty one valve to open er close at its reduced setpoint does initial opening.

not violate the design basis.

8 3/4 4-1 FERMI - Lpt!T 2

E LIST OF F!a g g PAGE FIGURE 3.1.5-1

$001UN PENTADDRATE VOLLBE/ CONCENTRATION REQUIREMENTS................................... 3/4 1-21 3.2.1-1 EXI M AVERAGE PLANAR LINEAR NEAT GENERATION RATE ( E PLMGR) VS. AVERAGE PLANAR EXPOSURE, i

INITIAL CORE FUEL TYPE SCR183................. 3/4 2-2 3.2.1-2 EXIM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (EPLHGR) VS. AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE SCR233.................. 3/4 2-3 N

3.2.1-3 '.

MAXI M AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE,

.I,NITIAL CORE FUEL TYPE SCR071.................. 3/4 2-4 3.2.3-1 MINI M CRITICAL POWER RATIO (MCPR) VERSUS Y AT RATED FLOW................................ 3/4 2-8 3.2.3-2 FLOW CORRECTION (K ) FACTOR.................... 3/4 2-9 g

q. y. f. v-/

THERNAL POWER VS CORE FLOW..................... 3/4 4-

' 1.1-1

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. I MAC AC m, S.SURE SSu tat TEN,ERATUREnSSu SSU...................... m 4-21

3. 4... r1 VS.

3. 4.10 ~/

THELMM fcJEt \\/S COAE Flew 3 M +~

4.7.5-1 SAMPLE PLAN 2) FOR SNUBBER FUNCT10EL TEST..... 3/4 7-21 B 3/4 3-1 REACTOR VESSEL WATER LEVEL..................... 8 3/4 37 8 3/4.4.6-1 FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE........................ B 3/4 4-7 LOCAL POOL TEWERATURE LINIT................... S 3/4 6-5 B 3/4.6.2-1 5 3/4.7.3-1 ARRANGE. NT OF SHDRE BARRIER SURVEY POINTS...... B 3/4 7-6 5.1.1-1 EXCLUSION AREA................................. 5-2 5.1.2-1 LOW POPULATION ZONE............................ 5-3 5.1.3-1 E P DEFINING UNRESTRICTED AREAS AW SITE 3002 ARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS...................................... 5-4

(

l FERNI - UNIT 2 xxi Amendment No.30

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Insert 4 Se NCFR surves illustrated la Figures 3.3.3-1 thra 3.3 5-1B were derived as described above for the following assumed operating r senditions:-

- curve A '- ScrR limit with turbine bypass system, moisture in servios and ccc separator reheater systess (control emil cere)less than or equalaparational made (A2 rods, Al to notch shallows inserted ition 36, all peripheral reds, and all rods f

He to position 46) inserted in the core.

operating domain includes the 1004 power / flow region and estended load line region with 2004 power and 5

reduond flow.

~

curve 3 - Herm limit with the turbine bypass system, moisture

'ereparator reheater systems in servios and non-CCC operational ande (

control rod inserted in the core).

The operat domain includes the 100%

power / flow region and the extended load line region with 1004 power and reduced flow.

Curve C - scPR limit for either Occ or non-ccc operational modes with either the main turbine bypass system inoperative and the moisture separator reheater system available or the main turbine bypass system available and the moisture separator reheater system inoperable. The operating domain includes the 100%

power / flow region and the extended load line region with 1004 power with reduoed flow.

Curve D - NCPR limit for either Occ or non-CCC operational modes with the main turbine bypass system inoperative and the moisture separator reheater The operating domain includes system inoperable.the 1904 power / flow region and the extended loa line region with 1004 power and reduced flow.

Curve A provides the NCPR limit assuming operation above 25 percent RATED TEERNAL POWER with the turbine bypass system and moisture separator reheater in service. The curve was developed based upon the operatiny McPR limits for a rod withdrawal error transient (UFSAR, Sect:,on 15.4.2) for operating within the CCC control rod patterns and a Main Turbine Trip with Turbine Bypass Failure CCC control rods are A2 rods, transient (UFSAR, Section 15.2.3).

Al shallow rods (inserted less than or equal to notch position 36),

The all peripheral rods, and all rods inserted to position 46.

analysis of the Main Turbine Trip with Turbine Sypass Failure takes credit for the steam flow to the moisture separator reboater.

Curve 3 provides the MCPR limit assuming operation above the 25 percent RATED TNERNAL POWER with the turb me bypass system and

.........,,_i

=

o meisture esperater sehmeter system la saavios and son-CCC eentrol Non-ccc eestrol rods are all zoos o

- - sods laserted in the sere.

emoloding 42 reds,11 aballow reds (inserted less then or equal to i

I notek position ss), all peripheral reds, and all rods imported tothe operating IIcra hosed position es. Who serve was devel isnt

, Section 15.4. 2) limits for h red withdrawal arrer for any operating withdreval engeenos.

curve C des the IIcrR limit operation above the 25 M

EERIBUG. POWER with the no separator reheater W-:Me and turbine bypass system inoperable er the moisture rehmeter inoperable and the turbine bypass system

=

eg.e.

The serve uns developed based upon the operating NCPR separatar operable.

limits for several sembiastions of Feedveter controller Failure.

curve D provides the IIcrA limit ensuming operation above the. 25 s

RATED THEIDDG, DOWER with both the moisture separator The percent reheeter i~noperable aria the turbine bypass system inoperable.

curve was developed based upon the operating NCFR limits from the Feedwater controller Failure.

9 O

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9 9

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i ese111ations is ineressed. During transients, the operator any not have sufficient time to annually insert sentrol rods to altigate the

(

esoillations before they reach an = maaaleah1a segnitado. Therefore, the L

prompt action of annually M the plant innen Region & is entered is required to ensure protection of the BJCPR.

Resed on test and operating esperianos, the frequency of core ther$

bydraulio eso111ations is less in Region 3 then in Region A.

Becay ratios are expooted and predicted to be louer in this region since Region B oovers a lower power and higher flow range than Region A.

Also, the margin to the EJCPR will typically be larger in Region 5 than in Region A.

With more margin to BJCPR and a lower probability of eso111ations, emiting Region B by control rod insertion is A tified. Sousver, if oscillations are observed uhl.'s exiting Region B, the reactor will be amana11y scrammed.

The potential for oore thermal hydraulio oscillations to occur outside of Regions & and B is very small and therefore special requirements are not shoessary outside of these regions.

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1 I

C

I a

Insert G:

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l he inpact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted at power levels t up to 70% of MTED 'IEEINAL POiER if the M:PR fuel cladding safety limit is increased as noted by Specification 2.1.2.

APM scram and control rod block metpoints (or APM gains) are adjusted as noted in Tables 2.2.1-1 and MAPLHIR limits are decreased by the factor given in 3.3.6-2, respectively.A time period of 4 boars is allowed to make these Specification 3.2.1.

adjustments following the establishment of single loop operation since the MCPR operating need for single loop operation often cannot be anticipata3.

limits adjustments in Specification 3.2.3 for different plant operating situations are applicable to both single and two recirculation loop operatiort.

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

1 additionally, surve211ance on the pung speed of operating recirculation loop is inposed to exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below 30% THEINAL PO4ER or 50% rated recirculation loop flow is to prevent undue thermal stress on vessel nozzles, recirculation puup and vesset bottom head durmg I

a power or flow increase following extended operation in the single recirculation loop mode.

Insert H:

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be mamtama3 durmg two loop operation, contmued operation is permitted in a smgie recirculation loop mode.

Insert I:

Sudden equalization of a temperature difference >l45 F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

Requirements are inposed to prohibit idle loop startup above the 80% rod line to minimize the potential for initiating core thermal-4draulic instability.

o E

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SASES IEEE SECTION t

INSTRUMENTATION (Continued)

MDNITORING INSTRUMENTATION (Continued)

Meteorology cal Monitori ng Instrumentation....... 8 3/4 3-4 Remote Shutdown System Instrumentation and B 3/4 3 Controls.........................

Accident Monitoring Instrumentation.............

9 3/4 3-4 B 3/4 3-4 Source Range Monitors...........................

' Traversi ng b-Core Probe System................. 8 3/4 3-4 8 3/4 3-5 Chletine Detection System.......................

Fire % tection Instrumentation..................

8 3/4 3-5 8 3/4 3-5 Loose-Part Detection System.....................

Radioactive Liquid Effluent Monitoring 8 3/4 3-5 Instrumentation.................................

Radioactive Gaseous Effluent Monitoring 8 3/4 3-6 Instrumentation.................................

8 3/4 3-6 3/4.3.8

, TURBINE OVERSPEED PROTECTION SYSTEM.............

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEMS ACTUATION 8 3/4 3-6 INSTRUMENTATION.................................

3/4.4 REACTOR COOLANT SYSTEM 8 3/4 41 3/4.4.I' RECIRCULATION SYSTEM...........................

A 3/4 41 3/4.4.2 SAFETY / RELIEF VALVES............................

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 8 3/4 4-2 Leakage Detection Systems.......................

8 3/4 4-2 Ope rati onal Leakage.............................

8 3/4 4-2 3/4.4.4 CHEMISTRY.......................................

8 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY...............................

8 3/4 4-4 3/4.4.6 PRESSURE / TEMPERATURELIMITS.....................

8 3/4 4-5

. MAIN STEAM LINE ISOLATION VALVES................

3/4.4.7 B 3/4 4-5 3/4.4.8 STRUCTURAL INTEGRITY............................

B 3/4 4-5

(

RESIDUAL HEAT REM 0 VAL........................... rs 3/4 t/-

3/4.4.9 3/4 4e O CoCE TH E euk. HYOR4uut.STA90ory...

xiii FERM1 - UNIT 2 4

mw mmw-r

4 IEEE s

b

}

LIMITING COM ITIONS FOR OPERATION AM SURVEILLANCE REQUIREE NTS SECTION PAGE t

3/4.4 REACTOR COOLANT SYSTEM J/4.4.1 RECIRCULATION SYSTEM Recirculation Loops..................................

3/4 4-1 Jet Pumps............................................

3/4 4-4 Recirculation Pumps..................................

3/4 4-5 Idle Recirculation Loop Startup......................

3/4 4-6 3/4.4.2, SAFETY / RELIEF ALVES Safe,ty/ Relief Va1ves.................................

3/4 4-7

~

Safirt'y/ Relief Valves Low-Low Sst Function............

3/4 4-8 i

3/4 4.3 REACTOR' COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................

3/4 4-9 Operational Leakage..................................

3/4 4-10 3/4.4.4 CHEMISTRY............................................

3/4 4-13 u 4.5 S,EC1FIC AC1Iv!Tv....................................

3/4 4-1.

3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Cool ant Sys tes..............................

3/4 4-19 Reactor Steam Dome...................................

3/4 4-23 3/4.4.7 PRIN STEAM LINE ISOLATION VALVES..................... ~ 3/4 4-24 3/4.4.8 STRUCTURAL INTEGRITY.................................

.3/4 4-25 1.

3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown.........................................

3/4 4-26 I

Cold Shutdown........................................

3/4 4-28 5 4 M.to Ccote TNGRMhL M t'OR4ut tc. STA R t t t ry.....

5/4 ht~

l 1

3/4.5 EERGENCY CORE COOLING SYSTEMS

. j i

3/4.5.1 ECCS - 0PERATING.....................................

3/4 5-1 3/4.5.2 ECCS - SHUTD0WN......................................

3/4 5-6 3/4.5.3 SUPPRESSION CHASER..................................

3/4 5-8 i

FERMI - UNIT 2 vi

~_-

m

Of' 3,

e RgACTOR mnr auf SYSTEM

j. -

t 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY BWR cores typically operate with the presence of global neutron flux noise in a' stable mode which is due to randon boiling and flow noise. As the power / flow conditions are changed, along with other systes parameters (pressure, subcooling, power distribution, etc.) the thermal hydraulic /

reactor kinetic feedback nechanism can be enhanced such that randos par,turbatiens may_ result in sustained limit cycle or divergent oscillations in power and flow.

Two major [sodes of oscillations have been observed in BWRs..The first mode is the fundamental or core-wide oscillation mode in which the entire core oscillates in phase in a given axial plane. The second mode involves regional oscillation in which one half of the core oscillates 180 degrees out of phase with the other half. Studies have indicated that adequate margin to the Safety Limit Minimus Critical Power Ratio (SLMCPR) may not exist during' regional oscillations.

Regions A and B of Figure 3 4.10-1 represent the least stable conditions of the plant (high power / low flow). Region A and B are usually entered as the result of. a plant transient (for example, recirculation pump trips)'and therefore are generally not considered part of the normal operating domain. Since all stability events (including test experience) have occurred.in either Region A or B, these regions are avoided to minimize the possibility of encountering oscillations and potentially challenging the SLMCPR.. Therefore, intentional operation in, Regions A or B is not allowed. It is recognized that during certain abnormal conditions within the plant, it may become necessary to enter Region A or B for the purpose of protecting equipment which, were it to fail, could impact plar,t safety or for the purpose of protecting a safety or fuel operating limit.

In these cases, the appropriate actions for the region entered would be performed as required.

Most oscillations that have occurred during testing and operation have occurred at or above the 100% rod line with core flow near natural circulation. This behavior is consistent with analysis which predict reduced stability margin with increasing power or decreasing flow. As core flow is increased or power decreased, the probability of oscillations occurring will decrease. Region A of Figero 3.4.10-1 bounds the majority of the stability events and tests observed in GE BWRs. Since Region A represents the least stable region of the power / flow operating domain, the potential to rapidly encounter large magnitude core thermal hydraulic

= _

p._. -

4 9

esoillations is increased. During transients, the operator any not have amfficient time to manually insert control rods to altisate the e{

ese111ations before they reach an unacceptable angnitude. Therefore, the 1

prompt action of annually $oramming the plant uhen Region A is entered is required to ensure protection of the BJCPR.

Ensed on test and operating experience, the firequency of more therbkl hydraulic oscillations is less in Assion 3 than in Region A.

Decay ratios are empooted and predicted to be louer in this region staos magion B oovers a louer power and higher flow range than Region A.

Also, the margin to the RJCPR will typically be larger in Region B than in Region A.

With more margin to RJCPR and a lower probability of oscillations, exiting Region 8 by control rod insertion is,)ustified. Nousver, if oscillations are observed while exiting Region B, the reactor will be manually scrammed.

i The potential for core thermal hydraulic oscillations to occur outside of Regions A and B,isJvery small and therefore special requirements are not noosssary outside of these regions.

i O

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.