ML20078C937

Amend 61 to License NPF-3,modifying Tech Specs to Permit Operation for Cycle 4 & Incorporating Revised Reactor Protection Sys Instrumentation Trip Setpoints & Allowable Values

ML20078C937
Person / Time
Site:	Davis Besse
Issue date:	09/21/1983
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Toledo Edison Co, Cleveland Electric Illuminating Co
Shared Package
ML20078C940	List: ML20078C937 ML20078C951
References
NPF-03-A-061 NUDOCS 8309280360
Download: ML20078C937 (61)

Text

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J-UN:T50, STATES NUCLEAR REGULATORY COMMisslON

$.,,.8hf._jj j WASHWGToN, D. C. 20555 YW1/i

%  ?,/

THE TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Anendment No.61 License No. NPF-3 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by The Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated July 5,1983, complies with the standards and re of the Atomic Energy Act of 1954, as amended (the Act)quirements and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is easonable assura~nce (i) that the activities authorized by this amendment dan be conducted without endangering the health and safety of the public, and (ii) that such activities.will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements i

have been satisfied.

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e 8309280360 930921 PDR ADOCK 05000346 l

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2.

Accordingly, Facility Operating License No. NPF-3 is hereby acended as indicated below and by changes to the Technical Specifications as indicated in the attachment to this license amendment:

Revise paragraph 2.C.(2) to read as follows:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised tnrough Amendment No.61, are i

hereby incorporated in the license.

The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR,THE NUCLEAR REGULATORY COMMISSION

- (Al 6h E. Stolz, Chief Op ' ting Reactors Branch #4 vision of Licensing

Attachment:

Changes to the Technical Specifications Date of issuance: September 21, 1983 S

9

i ATTACHMENT TO LICENSE AMENDMENT NO. 61 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment nuser and contain vertical lines indicating the area of change. The corresponding overleaf pages are als.o provided to maintain document completeness.

Paces 2-2 3/4 1-36 2-3 3/4 1-37 2-5 3/_4 1-38 2-7 3/4 1-39 B2-2 3/4 1-40 B2-4 3/4 1-41 B2-5 3/4 1-42 B2-6 3/4 1-43 3/4 1-26 3/4 2-1 3/4 1-28 3/4 2-2 3/4 1-28a 3/4 2-2a 3/4 1-28b

'3/4 2-2b 3/4 1-28c 3/4 2-2c 3/4 1-28d 3/4 2-2d 3/4 1-29 3/4 2-3 3/4 1-29a 3/4 2-3a 3/4 1-29b 3/4 2-3b 3/4 1-29c 3/4 2-3c 3/4 1-29d 3/4 2-3d l

3/ 4 1 - 31 B3/4 1-2 s

3/4 1-34 B3/4 2-2 l

3/4 1-35 l

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFEI'f LIMITS i

REACTOR CORE i

2.1.1 The c:mesnation of the reactor coolant core outlet pressure and outlet tamoerature shall not exceed the safety limit shown in Figure 2.1-1.

f AppLICA8tt m : MODES 1 and 2.

i ACTICN:

I Whenever the point defined by the cambination of reactor coolant core outlet pressure and outlet tamperature has exceeded the safety limit, 4

be in HOT STAN08Y within one hour.

REAC~0R CORE 2.1.2 The cambination of reactor THERMAL POWER and AXIAL POWER IMBALANCE shall not exceed the safety limit shown in Figure 2.1-2 for the various comeinations of two, three and four reactor coolant pump operation, i

j AFSLICABIL A: MCCE 1.

AC TCN:

Whenever the point defded by the enenination of Reactor Coolant Systam flow, AXIAL POWER IMBALANCE and THERMAL PGdER has exceeded the appropriate safety limit, be in HOT STAND 8Y within one hour.

REAC~0R COOLANT SYS~EM 88LE53URE 2.1.3 The Reac ce Coolant Sys am pressure shall not exceed 2750 psig.

AP.ICA3 fLITY : MODES 1, 2, 3, 4 and 5.

AC*TCN:

M00E51 and 2 - Whenever the React.ar Coolant Sysan pressure has ex-caeced 2750 psig, be in HOT STANCBY with the Reac~cc, Coolant Systen pressure within its limit within one hour.

MODES 3 A

- Wherever the Reactor Coolant System pressure has and 5 exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 zinutes.

DAVIS-BESSE, UNIT 1 2-1

D D

Figure 2.1-1. Reactor Core Safety Limit 2500 2400 RC HIGH PRESSURE TRIP 618.2300

^

RC HIGh TEMPERATLRE 3

2200 7 gip 618,2124.6

$l.

ACCEPTABLE OPERATION RC PRESSURE TEMPERATURE TRIP 606.79, 1983.4 SAFETY t.lMIT 2000 RC LOT PRESSURE TRIP 1800 f

f I

f 1

I 590 600 610 620 630 540 Reactor Outlet Temperature. *F 4

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2-2 DAVIS-BESSE, VilIT 1 Amendment flo. 77, 23, JE,61

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s Figure 2.1-2. Reactor Core Safety Limit

f. R A TED TH ERM A L POWER h

4PW 120 (33,112)

(-31,112)

LIMIT 100 (W,100)

( -46,100)

(-31,89.24)

(33,89.24) 3 PUMP 8

( -a,n. 24) <

> ( w, n. 24)

ACCEPTA8LE

-60 OPERATION UNACCEPTABLE UNACCEPTABLE FOR SPECIFIED OPERATION OPERATION RC PUMP COMBINATION

. go

-20 t

i f

f f

-60

-40

-20 0

20 40 60 Axia l Power imbalance, f.

PUMPS OPERATING REACTOR COOLANT FLOW, GPM 4

387,200 3

290,l00 l

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l 2-3 DAVIS-BESSE, UtlIT 1

- Amendment :'o. 77, 73, 33, pg,61

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM SETPOINTS 2.2.1 The Reactor Protection System instroentation setpoints shall be set consistent with ne Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a Reactor Protection System instrumentation setpoint less conserv-ative than the value shown in the Allowable values column of Table 2.2-1, declare the channel inoperable and aoply the applicable ACTION statament recuirement of Specification 3.3.1.1 until the enannel is restored to OPERABLE status with its trip setpoint adjusted consistent with the T.'ip 5etpoint value.

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DAVIS-BE3!E, U'i!T 1 2-4 I

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Table 2.2-1. Reactor Protection System Instrumentation Trip Setpoints na l'1 Functional unit Trip setpoint Allowable values

• i 1.

Manual reactor trip Not applicable.

Not applicable.

2.

liigh flux

<104.94% of RATED TilERNAL POWER with

<104.94% of RATED TilERHAL POWER witir l

Tour pumps operating Tour pumps operatin98 (79.85% of RATED TilERHAL POWER with

<79.85% of RATED TilERHAL POWER with three pimps operating Three pumps operatingI l

3 ItC high temperature 1618*F

_618'FI 4.

Flux -- aflux/fl ow(1)

Trip setpoint not to exc'eed the lim-Allowahle values not to exceed the it line of Figure 2.2-1 limit line of Figure 2.2-13 n

S.

ItC low pressure (I) 11983.4 psig i

11983.4 psig*

>1983.4 psig**

6 ItC high pressure 12300 psig f2300.0psig*

12300.0psig**

7.

ItC pressure-temperature (l) >(12.60 Tout F - 5662.2) psig

>(12.60 Tout F - 5662.2) psigI l

8.

llinh fluqnumber of RC

<55.1% of RATED TilERHAL POWER with

<55.1% of RATED THEllHAL POWER wi th pumps ont J k

one pump operating in each loop one pump operating in each loopI E.

(0.0% of RATED TilERHAL POWER with (0.0% of RATED TilERHAL POWER witti j

two pumps operating in one loop and Two pumps operating in one loop and no planps operating in the other loop no pimps operating in the other loopf Il (0.0%of RATED TilEltHAL POWER with no

<0.0% of RATED THERNAL POWER with no pimips operating or only one pump op-pumps operating or only one pump op-erating y

erating#

8 9.

Containmenit pressure high

<4 psig y

i 14 psigI

,t O

b.

4

M,,,.

-~~

E

5 T'

h Table 2.2-1.

(Cont'd)

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Trip may be manually bypassed when'RCS pressure $1820 psig by actuating shutdown bypass provided that:

U a.

The high flux trip setrotut is 55% of RATE!) TilERMAl. POWER.

~

b.

The shutdown bypass high pressure trip setpoint of s1820 psig is imposed.

c.

The shutdown bypass is removed when RCS pressure >l820 psig.

• Allowable value for CllANNEL FUNCr10NAI. TEST.

t

• • Allowable value for CilANNEL cal.IBRATION.

1. Allowable value for CilANNEL FUNCTIONAL TEST and CilANNEL cal.IBRAT10;4.

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Figure 2.2-1. Trip Setpoint for Flux -. Flux / Flow i

Curve shows trip.setpoint for a 257 flow reduction for tnree pump operation (290,l00 gpm). The actual setpoint wi.nl be. directly proportional to the actual flow with three pumps.

7, RATED THERMAL POWER UNACCEPTABLE

- 120 OPERATION UNACCEPTABLE OPERATION

(-14,106.9)

( 17,106.9)

N = -l. 8576 M =0 9288 s

l 2

r PUMP 100 LIMIT l

(-32,90.182)

I (26,90.182)

)

g(-14,79 85)

( g7,79,35 )

l 3 PUNP l

l I

LINIT l

(-32,83.132)

. 60 l

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lACCEPTABLEOPERATIONFORl l SPECIFIED RC PUNP COMB-l l lMATION

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- 20 l

y!

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'Lj 1

1 % I,,

=

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-60

-40

-20 0

20 40 60 Axial Power imbalance, f.

I DAVIS-GESSE, UtiIT,1 2-7 Amendment i:o. 77, 75, 33, M 61

Figure. 2.2 2 A.11ovable value for nu-A na/now D" O s

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-S Amendnent No. JP,Xf,7f,45

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2.1 SAF Tf L.'M.I 3 3ASEI 2.1.1 and 2.1.2 REAC CR CORE The restrie:1ons of nis safety limit ;rsvent overneating cf the 'uel cladding and possible cladding perf: ration wnica would result in the

-elease of fission products :s the rescur coolant. Overneating of the fuel cladcing is prevented by restricting fuel operation to within tne transfer coefficient is large and nucleata Sciling regime wnert the nea

ne cladding sur acs *,ammeraurt is sligntly acove :ne =olan: satura:1cn d

tamvers:urt.

Omeration aseve the uceer boundary of the nucleate boiling regime would result in excessive cladding tammeratures because of :ne onset of 0

nucleats boiling (DNB) and ne resultant sharp recuction

ecarturt in nes: transfer = efficient. DNS is not a directly measurable parsmeter

uring Operation anc :nerefore THERMAL PCWER and Reactor C clan Temper-anc P-Tssure nave been related to DNS :nreugn the S&W-2 DNS a:Urt

=r-tiation. The CNB cor-slation has been developed :s predic One DNS l; #1u anc ne 10:st;on of CNB for axially unifem and non-unifem nea:

The local DHB nea flux ratio, CNSR, defined as :ne

l ' lux :ist-ibu; ens.

l P2-e Of

ne neat flux cat would cause CNS-at a particular care loca:icn

' = ne local nes; flux, is indicative of :ne margin = DME.

' he =1nimum value :1 ne CNBR during s-ancy state coeration, nomal

. =e-2:icnal : ansients, and anticipatad :ransients is limited to 1.20.

his value =r es==nes it a 95 mercent precacility at a 95 percent i

=nH:enen level na: CNB will not cc:ur and is esosen as an accrecriata targ n :: 2N51:r all coerating ecnditions.

he =ne :resentad in Figurt 2.1-1 recrasents the esnditions at wnica a finimu= IN5R Of 1.20 is precic.ad f r One maximum possible themal power 287, 200 GFM, wnich is llc: 01

. 2 nen :ne atac.:t colan: ' low isesign ".cw esta for f:ur opersting renc=r c:

This :urve is l tasac :n ne foll: wing he: =annel fac=rs w1:n potential fuel censifi-t

s:::n an: 'uel rec Ocwing effec s:

. = 2.561

.

• I. 71 ;

• 1.50

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ower :esting fac=rs are One most rest-ictive he :est;n li=-:full :ower f r ce range # rem all =ntni reds fully l

si:ulatec a:

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wit.c-swn = min =um ailemacle =ntrol red winorswal, and fem :ne

=rt 'NBR :esign : asis.

s Anenc=en: No..[ 3 3 321 OA V 5 -5 EI~ I, :.*N T 1 l

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jiSAFETYLIMITS

'i l3ASES

The reactor trip enveloce appears to approach the safety limits more closely

'than it actually does because the reactor trip pressures are measured at a lo-cation where the indicated pressure is about 30 psi less than core outlet pressure, providing a more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densificaticn and coten-

, ;;ial fuel rod bow. '

'1 1.

The 1.30 DNBR limit procuced by a nuclear power peaking factor of FO*

2.56 or the comoination of the radial peak, axial ceak, and position of the axial peak that yields no less than a 1.30 DNBR.

2.

The combination of radial and axial peak that causes central fuel meltino at the hot spot.

The limits are 20.4 kW/ft for batenes 10, 23, 4 anc 5A l

I and 20.5 kW/ft for batches 5B and 6.

Power peaking is not a directly observable cuantity a'd therefore limits nave n

' t oeen established on the basis of the reactor power imoalance procuced by :ne

' cower peaking.

The specified fl ow rates for curves 1 ano 2 of Figure 2.1-2 correscond to tne l expected minimum flow rates with four pumos and tnree pumos, respectively.

lThe curve af Figure 2.1-1 is the most restrictive of all oossible reactor

, coolant pumo-maximum thermal power combinations shown in BASES Figure 2.1.

The curves of BASES Figure 2.1 recresent tne conditions at wnicn a minimum DNBR of 1.30 is predicted at the maximum possible tnermal power for :ne num-cer of reactor coolant cumes in operation or tne local cuality at tne point of minimum DNBR is ecual to +22%, whichever condition is more restrictive.

ll These curves include the potential effects of fuel rod bow and fuel censifica-Il tion.

!iThe DNBR as calculated by the B&W-2 Df8 correlation continually increases

from coint of minimum Of8R, so tnat the exit OPER is always nigner.

Extraco-

!ilation of the correlation beyond its ouolisned cuality rance of -225 is justi-

' fied on the basis of exoerimental data.

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3 2-2 DAVIS-BESSE, UtlIT 1 Amendmer.: io. 77, 33, 4E41

SA C LIMI 5 3ASE3 For og curve of SASIS Figurt 2.1, a pressure-tannerature point acove and :s me left of :ne curve would result in a DNBR greater than 1.20 cr a lor.al cuality at ne point of minimum 7,N8R less can *22

• for tr.a = articular reactor ecolant puso situation. The 1.30 DNBR l curve for three pump operation is more restrictive tnan any other reactor coolant pump situation because any pressure /temaerature point above and to :ne left of the three pump curve will be above and to the left of the four pumo curve.

j 2.1. 3 ;EAC OR C CLANT SYS3 79EISURE l

"he ts:M :icn Of :nis Safety Li=it tretects tne integrity of the teact:r Occian: Systam f m everpressuri:ation anc meracy prevents :ne i

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tiease Of ndienuclices =ntained in ne reac..r c:clant fr:m reacning ce =nts:nment a=cs:ners.

he tac.:r :rtssure vessel and Ortssuri:ar are casigned to Section i ~:: Of =e GiE Sciler anc Pressurt 'iessal 0:ce wnica set ni.s a maxi =um

! - snsien: Ortssurs =f 11C"., 2750 :sig, Of esign artssure. The lanc :r

'I ::::an: Sys.am si:tng, valves sne 'i:-ings, are designed u ANS~ S 31.7,

155 E:1: en, wn,= :e-.t:s a maxi =um transien: ;rtssurt W 11C:, 2750

si;, :f==:ener.: des ;n :ressure. The Safety Lim 1: of 2750 psig is

-l ne ef:rt =nsistan:

wiu =e :esign : 1: aria and asscciated c:ce scui sments.

he en irs teact:r ::lan: System is hycrt:estad at 3125 : ig,125t

emens: ata integrity prior u initial coerstion.

' esign :rtssure, ::

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• 5.i EIII, Ji 'T

• 3 I-3 Amenc:Tient No. R,72,45

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! ! 2. 2.

LIMITING SA~ETY SYSTEM SETTINGS

!l

'IBASES I

2.2.1.

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The reactor protection system instrumentation trip setpoints specified in Table 2.2-1 are the values at which the reactor trips are set for each param-eter.

The trip setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

The snutdown bypass p.ovices for bypassing certain functions of the reactor il orotection system in order to permit control rod drive tests, zero cower :HYS-l ICS TESTS and certain startuo and snutdown procedures.

The purpose of the shutdown bypass high pressure trip is to prevent normal operation with snut-down bypass activated.

This high pressure trip setcoint is lower than the normal low pressure trip setooint so that the reactor must be tripoed before the bypass is initiated.

The high flux trip setooint of <5.0% prevents any significant reactor power fran being produced.

Sufficient natural circula-l tion would be available to remove 5.0% of RATED THERMAL POWER if none of :ne reactor coolant pumps were operating.

Manual Reactor Trio I,

I The manual reactor trio is a redundEnt channel to the automatic reactor orotec-i tion system. instrumentation channels and provices manual reactor trio cacaoil-

lity, d

!' Hion Flux i'

ii A high flux trio at high power level (neutron flux) provices reactor core oro-

tection against reactivity excursions wnich are too raoid :o me protected by l } temperature and pressure protective circuitry.

Il1 Ouring normal station operation, reactar trip is initiated wnen the reactor i power level reacnes 104.94% of rated power.

Due to transient oversnoot, hea

'i balance, and instrument errors, the maximum actual power a: wnien a trie j ; would be actuated could be 112%, whicn was usec in the safety analysis.

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t-5' 0 AVIS-BESSE., UNIT 1 Amendment !!o. A3,61

t ' LIMITING SAFETY SYSTEM SETTINGS tl l BASES i

, RC Hien Temocrature The RC high temperature trip <61S*F prevents the reactor outl et temperature from exceeding the design lidits and acts as a backup trip for all cower ex-cursion transients.

! I Flux -- 1 Flux / Flow il

!l The power level trip setpoint produced by the reactor coolant system n ow is

l based on a flux-to-flow ratio which has been established to accommodate flow ll decreasing transients from high power where protection is not proviced by tne j i high flux / number of re&ctor coolant pumos on trips.

li: t The power level trip setpoint produced by the cower-to-fl ow rstio orovices I both high power level and low flow protection in the event the reactor cower i i level increases or the reactor coolant fl ow rate decreases.

The power level 6 ! setpoint produced by the power-to-flow ratio provides overpower DNB protec-ll tion for all modes of pump ooeration. For every flow rate there is a maximum l' oermissible power level, and for every power level there is a minimum permis-l sible l ow fl ow rate.

Examples of typical power level and l ow fl ow rate com-

l Dinations for the pump situations of Table 2.2-1 that would result in a trip

' are as follows

i i

1.

Trip would occur when four reactor coolant pumos are coe.3:ing if oower i

is 106.9% and reactor coolant flow rate is 100% of full fl ow rate, o r i

1 fl ow rate is 93.5% of full fl ow rate and power level is 10C%.

f

! j 2.

Trip would occur when three reactor coolant cumos are operatino if oower is 79.9% and reactor coolant fl ow rate is 74.7% of full fl ow rate, or ti i l flow rate is 70.2% of full n ow rate and power is 75%.

ti For safety cal culations the instrumentation errors for tne power level were eused.

Full flow rate in the aoove two examoles is defined as tne M ow calcu-

- lated by the heat balance at 100% cower,

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DAVIS-BESSE, U!IT 1 Amendment l'o. 79, 33, fE,6-I

llLIMITINGSAFETYSYSTEMSETTINGS l

li3ASES The AXIAL POWER D98ALANCE boundaries are established in order to orevent reac-tor thermal limits from being exceeded.

These thermal limits are either power peaking kW/ft limits or DNBR lindts.

The AXIAL POWER IMSALAfCE reduces the power level trio produced by a flux-to-flow ratio such that the bounda-ries of Figure 2.2-1 are produced.

f RC Pressure - Lew. Hicn. and Pressure Temoerature The high and low trios are provided to limit the pressure range in wnien reac-

, ' tor coeration i s pe rmi tted.

i-l! During a slow reactivity insertion startuo accident from low power or a slow li reactivity insertion from high cower, the RC high pressure setpoint is reached before the high flux trio setooint.

The trip setooint for RC hign pressure, 2300 asig, has been established to maintain the systen pressure be-low the safety limit, 2750 psig, for any aesign transient.

The RC hian cres-i

. l sure trip is backed up by the pressurizer code safety valves for RCS over li pressure protection, and is therefore set lower than the set oressure for 2525 psig. The RC high pressure trip also backs up the high l;l these valves,1 flux trip.

n ll The RC low pressure,1983.4 psig, and RC pressure-temoerature (12.60 t

5662.2) osig, trip setpoints have been established to maintain the DN3 ra

io ou i, greater than or equal to 1.30 for those design accidents that result in a i j oressure reduction.

It also prevents reactor coeration at pressures celow

} the valid range of DNB correlation limits, protectino against DNB.

s' l Hich Flux /Numoer of Reactor Coolant Pumos On ll In conjunction with the flux - aflux/ flow trio the high flux /nuccer of reac-ij :or coolant pumos on trio prevents the minimum core DNBR from cecreasing

', below 1.30 by tripping the reactor due to tne loss of reactor coolan:

ji cumo(s).

The pumo monitors also restrict the cower level for the nuccer of

j oumos in coeration.

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'l 32-0 DAVIS-BESSE, UNIT 1 Anehd cnt Mo. 22, /3, 32,61

6 O

REACTIVITY CONTROL SYSTDG Sar??Y R00 INST: TION L!w:7 L !* QN-CONDITION FOR OPC'.tATION 3.1.3.5 All safety rods shall be fully withdrawn.

ADDLICAEILITY:

l' and 2**.

ACTION:

Witn a maximum of one safety red not fully withdrawn, except for sur-veillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a.

Fully withdraw the red or Declare the rod to be inopertole and apply Specification b.

3.1.3.1.

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_ '!!!LLAN~I REOUTREME WS SUR a.1.3.5 Eacn safety red shall be detemined to be fully withdrawn:

Witnin 15 minu1!as prior to withdrawal of any regulating red a.

curing an a :roacn t: reacter criticality.

b.

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

'See 5:e:g. Test Exception 3.10.1 and 3.10.2.

8With K,ff > 1.0.

1 DAVIS-BE5SE. UN T 1 3/4125

'. REACTIVITY CONTROL SYSTE.%

REGULATING ROD INSERTION LIMITS r!

• LIMITING CONDITION FOR OPERATION If l13.1.3.6 The regulating rod groups shall be limited in pnysical insertion as

lshown on Figures 3.1-2a, 3.1-25, 3.1-2c, 3.1-3a, 3.1-3b and 3.1-3c.

• i

' APPLICABILITY:

MODES l' and 2**.

ii

', : ACTION :

• ! jWith the regulating rod groups inserted beyond the above insertion limits (in ita region other than acceptable coeration), or with any group secuence or over-

. lao outside the specified limits, except for surveillance testing oursuant to 1lSoecification4.1.3.1.2,either:

!\\

aja.

Restore the regulating groups to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or it i;b.

Reduce THERMAL POWER to less than or ecual to that fraction of RATED THER-MAL POWER which~is allowed by the rod group cosition using the aoove fig-ures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or ti'.c.

Be in at least HOT STANDBY within 6' hours.

iNOTE

If in unacceotable region, also see Section 3/4.1.1.1.

i.

ti i

t g I l

1. :

e i'

V l

I i

i l

i I

t 4 6

!'See icecial Test Excections 3.10.' and 3.10.2.

I f

.{witn'<,,

,1 I

!JAV:5-5E55E. 'JNIT 1 3/: 1-26 AmenJnent r:o. 77, 33, gy, AE, A7,61

. k,

l REAC TVtw C3TRCL SYS-m!

R GdLATING 200 tMSERTicM U MIT3, SURVE!!.LANC! RE0tJtRE=.3 U 4.1.3.5 ne position of enca regulating grous saall be detanrined u be witnin ce inser:1cn, sequenca and overlas liarf u a: leas: enca every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> excast unen:

a.

De regulating md insertion limit alarm is irrecerstle. :nen verify me greums :s de witnin me inser:1cn limits a: least onca per 4 Mours; 3.

De c:ntml red dHve secuence alarm is incoerable, then veMfy me gr:uss a be wicin ce sequenca anc overiam limits ac least enca par 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

I ll ll ll

'l

j I

i lI l

.I ll 'Av:5.GIIII, ;3r 1 3/4 1 27

c, Figure 3.1-2a.

Regulating Group Position Limits, O to 24 4 10. -0 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 m

E m

N (246,102 (275,802 Power Level (300,102) g 100 Cutoff = 1001 3 y) g,

HAR0lN

>(268,92) g LlHIT a0 i(265,a0) n.

J I

l ff, f

UNACCEPTABLE 60 a

OPERAll0N D

(Is5,50) d (225,50) s v

7' eg 40 ACCEPIABLE U

h OPERATION I

E

[

20 (112,15) g a

r

,2.4)

. (100.01 i

8 N;-

O i5 200 300 k

RodIndex(% Withdrawn)

s" GR S i I

3 y

0 75 100 O

GR 6g i

I

.I 0

25 75 100 N

GR 71 1

I g

0 25 100 g

e

.' g

u 3E c:

,k figure 3.1-2b.

Regulating Group Position Limits, 24 + 10, -0 to 150 + 10 EFPO, Four RC Pumps - Davis-Besse 1, Cycle 4 Ei A

(270,102) iOO Power level

( 2816,10 2)

<r~

o(300,IO2)

Cutoff = 100%

(262,92)

SHUIDOWN m

15 NARGIN LlHIT (261,80)

El c

!.h, f5 U[a UNACCEPTABLE OPERAll0N

' g 60

[

O (185,50) g (225,50) h'

'$ 810 l'

E ACCEPIABLE OPERATION y,

/

$ 20 (1 2.iS) o,~,

(

2. 11 )

(100,0) n.

0 g

i 8

0 100 200 300 A

RodIndex(% Withdrawn) f GR 5 I I

I O

75 100 0

GR 6 l l

l 1

y 0

25 75 iOO

%w GR 7 I I

.I 0

25 100 m

u@

lIUure 3.1-2c.

Regulating Group Position Limits Af ter 150 +10 EFPD, G

Four RC Pumps - Davis-Besse 1, Cycle 4 s'n g

(210,102)

~

(266,102)

O(300,102) il 100 -

Power ievel H

Cutofi = 1001 (262.92)

SilulDOWN

-(

, MARGIN 2 80 L8HII (255,80) d OPERAll0N RESIRICIED 0PE I N

/

8-c2 60 l

l'8 (215,50)

(225,50)

<a o

7.

..g 40 g

/>

e El E.'

ACCEPIABLE OPERAllok y, 20 3

E (1888,8 5) i 9

g.

o(

,2.3)

A_.100.8) i

=

0 t60 200 300 E+

IIad in6ct (f. Withdrawn)

I 1

GR S d

15 100

\\

GR 6 l l

l l

5.

O 25 75 100 9

GR 7 I I

I O

25 100 l

I

l

=

l I

I l

1 1

This page left intentionally blank.

DAVIS-BESSE, UNIT 1 3/4 1-28c Amendment No. M,61

This page left intentionally blank.

DAVIS-BESSE, UNIT 1 3/4 1-28d Amendment No. 45, 61

E

^

Figure 3.1-3a.

Regulating Group Position Limits, O te 24 + 10, -0 g

EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 N

y, c;

100 ;-

-i W

o 80 (246,77 (275.77) 0 (300,77) a SilulDOWN

_j "x

HARGIN (268,69.5) 4++8 Ei LlHIT j>(265,00.5) o 60 UNACCEPTABLE l'8 OPERATION

+4

'i

/

e

.g 40 (185,38 (225,38)

T a

's u

j 20 (l12,i1.75' ACCEPTABLE OPERAIION gy

,. H,.a) a (800 0) i i

k 0

100 200 300 il Rodindex(%Withdrawii)

N GR S l

l g

0 75 100 0

GR 6 l l

l l

~

w 0

25 75 100 m

GR 71 1

I D

0 25 100 0

Uy f igure 3.1-3b.

Itegulating Group Position Limits, 24 + 10

-0 to 150 j 10 1

4 EFPD, Three RC Pumps - Davis-Desse.1, Cycle 4

[2n 100 -

E S

g (2116, 77) (270,77) 30 c_

S SilblDOWN O(

}

M (262,69.5) r/

WRGIN y

LINIT s.s f

60 (261,60.5) a l'l;)

UNACCEP TABLE

{

d OPERATION 11 0 (185,38 (225,34)

C I

Y>

a.

S.

k, 20 (112,11.75)

ACCEPIABLE 01'LRAll0N t

,2.3)

I U'N I

1 O

w

$ iT 0

100 200 300

'{

Rodindex(f. Withdrawn) g GR 5 i

[

i

,3 0

15 100

=

GR 6I i

1 1

0 25 15 IM

,4" GR 7 I I

I to o

25 800

E G

4, Figure 3.1-3c.

Regulating Group Position Limits After 150 +10 EFPD, g

Three RC Pumps - Davis-Besse 1 Cycle 4 E

g 100 4

a

^m (266,77)(270,77) 80 n.

O(300,77)

SHUIDOWN f

NARGIN (262,69.5) i LlHIT j

60 (255,60.5)

UNACCEPTABLE kE OPERATlilN w

OPERAT10N RESTRICTED u

o l

{

g 40 (21,5,38)

(225,38)

E E.

g 20 2

(148,11.75)

ACCEPTABLE OPERATION 95 0 ('

' '2 y

0 100 200 300 Rod index (% WI thdrawn) 2 P

GR 5 1

I y

0 75 100 u

GR 6 l

l

~~)

O 25 75 100

.N GR 7 l l

[

g 0

25 800

This page left intentionally blank.

l DWIS-BESSE, UNIT 1 3/4 1-29c Amendment No. 77, 33, 6,61 l

l

'This page left intentionally blank.

DAVIS-BESSE, UNIT 1 3/4 1-29d Amendmt t No. AS,61

REACTIVITY CONTROL SYSTIMS RCD PROGRM LIMITING CONDITION FOR OPERATTON 3.1.3.7 Each control rod (safety, regulating and APSR) shall be pro-grassied to operate in the core position and rod group specified in Figure 3.1-4 l

AF8tICA8!LITY: MODES l' and 2*.

AC* ION:

With any contnl red not programed to coerate as specified above, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

SURVE!LLANCE REQUIREMENTS 4.1.3.7 E

a.

Eaca control rod shall be demonstrated to be prograzed to operata in the specified core position and rod group by:

1.

Selection and actuatien fr:st the control room and verifi-cation of movement of the proper rod as indicated by both

ne ansclute and relative position indicators:

a)

For all contnl rods, after the control rod drive patches are locked suesecuent to test, reprogranaring or maintenance within the panels.

b)

For specifically affected individual rods, following main'.enance, test, reconnection or modification of power or instrumentation caales fres :ne control red drive control system to the control rod drive.

2.

Verifying that eacn cable that has been disconnected has been properly matened and reconnec.ed to :ne specified control rod drive.

b.

At least once esca 7 days, verify that the control red drive pa::n panels ars locked.

'See spec:ai Tes: heestions 3.10.1 and 3.10.2.

2AV!S-SESSE. UNIT 1 3/4 1-30 Amendment.1o.11

t i

Figure 3.1-4.

Control Rod Core Locations and Group Assignments - Davis-Besse 1, Cycle 4 1

1 A

3 4

7 4

C 1

5 5

1 i

0 7

8 5

S 7

E 1

5 5

1 F

4 8

3 7

3 3

4 G

5 2

2 6

~I H

7 5

7 3

7 5

7

(

E.

2

--'2 6

1 8

3 7

3 3

A o

9 1

5 5

1 N

7 8

5 S

7 0

1 5

5 1

1 P

4

'7 4

R I

I 1

2 3

1 5

5 7

3 3 l 10 11 12 12 14 15 GROUP NO. OF 005

UNCTIONS I

3 SAFETY 2

3AIIII X

GROUP NU.9BER 3

5 SAFETY a

3 SAFETY 5

3 CONTROL 5

3

NTROL l

7

0NTROL 3

1

!?SRs TUTIL = 31 DAVIS-BESSE, Uf;IT l 3/4 1-31 Amendrient fio. II, 32, 13, 61

INTEiTICNAJ.LY L.=.7 4. LANK l

l

~

.o

..-:: a:.

p 1.v-m, en=mn:.4c..,.

~

REACTIVITY CONTRCL SYSTEwS IENON REAC*IVITY LIMITING CCN0! TION FOR OPERATION 3.1.3.8 THE.WL POWER shall not be increased above the power level cutoff specified in Figure 3.12 unless one of the following concitions is satisfied:

a.

Xenon react,1vity is witain 10 percent of the equilibrium value for RATED THERMAL PCWER and is approaching stability, or b.

THERMAL POWER has been within a range of 87 :s 92 percent of RAU.D THERMAL PCWER for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluole poison control moce, excluding xenon free start-ups.

APD'.!CABILITY: MCDE 1.

ACTION:

Wita :ne recuirements of the above specification not satisfied, recuce THERMAL PCWER :s less taan or equal :s ne power level cu.:ff witain 15 minutes.

SURVIILLANCE RECUIREMENTS A.l.3.3 Xenon reactivity shall be datamined to be within 10% of the equilibrium value for RATED THERMAL PCWER and to be approacning s. ability or it snall be detamined that the THERMAL POWER has been in :ne range :f 37 to 925 of RATED THERNAL POWER for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, prior to increasing THERMAL POWER above :ne power level cE: ff.

DAVIS.BEESE, UNIT 1 3/4 1-33 1

' REACTIVITY CONTROL SYSTEMS AXIAL PCWER SHAPING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.9 The axial power shaping rod group shall be limited in physical in-sertion as shown on Figures 3.1-Sa, 3.1-5b, 3.1-5c, 3.1-5d, 3.1-Se and 3.1-5f.

APPLICABILITY: MODES 1 and 2*.

ACTION:

With the axial power shaping rod group outside the above insertion limits, either:

a.

Restore the axial power shaping red group to within the limits within 2 i

hours, or b.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures.,within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or c.

Se in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SUR'lEILLANCE REQUIREMENTS 4.1.3.9 The position of the axial power shaping red group shall be deter-mined to be within the insertion limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except when the axial power shaping rod insertion limit alarm is incoerable, tnen verify the group to be within the insertion limit 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

} 'With k,ff 2 1.0.

I i

0 AVIS-3 ESSE, UNIT 1 3/4 1-3 Amendment ;io. 33, /3, #E,61

~

~

rigure 3.1-Sa.

APSR Position Limits, O to 24 + 10, -0, EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 (8,102)

(38,IO2) 100 C

O (8,92)

(38,92)

RESTRICTED REGION 2

w

! 80 Cr (0.80)

( 42,80)

NaM o 60 W

PERMISSIBLE 5

OPERATING REGION (iOO50) o fn E

5 b

e g20 c-a 0

e i

e i

i i

i i

I O

10 20 30 50 60 70 80 90 iOO APSRPosition(f.Witnerawn)

~

o DAVI5-BESSE, tmIT 1 3/4 1-35 Amendment No. 33, $5,61

Figure 3.1-5b.

APSR Position Limits, 24 + 10. -0 to 150+10 EFPD, Four RC Pumps - Davis-Besse 1 Cycle 4 (8,102)

(38,IO2) 100 C

9 (8,92)

(38,92) 80Cr (0,80)

(42,80)

RESTRICTED a

REGION f5 iE 60 W

(iOO,50) e t

PERMISSIBLE y

y OPERATING REGION ab

=

1.

h 20 Cl-0 f

f t

i 0

10 20 30 40 50 60 70 80 90 100 APSR Position (". Withdrawn) t DAVIS-BESSE, UllIT 1 3/4 1-35 Amendment flo. 33 A3.15, 61

Figure 3.1'-Sc. APSR Position Limits After 150 =10 EFPD, I

Four RC Pumps - Davis-Besse 1, Cycle 4 (8,102)

(38,iO2) 100 C

O (8,92)

( 38,9 2)

E5 80 (

(0,80)

(4,80) c c.

RESTRICTED REGION 60 U

PERMISSIBLE 5

OPERATING REGION t

(ico,50) y w

8 bb y

20 2

0 e

t t

t i

t i

0 10 20 2

M 50 60 70 80 90 100 APSR Position (7, Withdrawn)

DAVIS-BESSE,'U' LIT 1 3/4 1-37 Am'endment tio. 33, f), 61

Figure 3.1 - 5d.

AP'SR Positi n Limits, O to 24 + 10

-0 EFPD, Three RC P6mps - Davis-Besse 1, Cycle 4 120 g 100 W

E a

80 -(8.77)

(38,77)

M

[

(8,69. 5)

(38,69.5) w 3

60 ( (0,60. 5)

('42,60. 5)

RESTRICTED REGION-T GJ E

40 PERMI SSI BLE E

OPERATING REGf0N 5

(iOO,38) t 5

20 Q

f f

I t

I t

i I

f f

C 10 20 30 40 50 60

'70 80 90 100 APSR Posi tion (7, Wi thdrawn)

DAVIS-BESSE, UtiIT 1 3/4 1-38 Amendment fio. 33, /2, f),61 I

O Figure 3.1-5e.

APSR Position Limits, 24 + 10, -0 to 150+10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 120 g iOO

-(8.77)

(38,77) 80 RESTRICTED (8,69.5)

(38,69.5)

REGION g

(0.60.5)

(42.60.5) 60(

D eoj 40 m

[

PERMISSI BLE (100*38)

OPERATING REGION g

s s

20 0

I t

t t

0 10

?C 3C 40 50 60 70 80 90 100 APSR Position (f. Withdrawn)

DAVIS-BESSE, UNIT 1 3/4 l-39 Amendment No. 52, 45,61

Figure 3.lv5f.

APSR Position Limits After 150 :10 EFPD, 7hree RC Pumps - Davis-Besse 1, Cycle 4 120 '

E iOO W

2 80

. (8,77)

(38,77)

RESTRICTED REGION (8,69.5)

(38,69.5) c W

E 60 (

(0,60.5) 4,60.5) e T

8g u

c 5

PERMISSISLE (100.38) l OPERATING REGION a.

20 0

I

+

t i

i i

0 10 20 30 13 0 50 60 70 80 90 100 APSR Position (f. Wi tacrawn)

~

DAVIS-3 ESSE, UNIT 1 3/4 1-40 Amendment No. fE, 61

O O

k s

This page left intentionally blank.

DAVIS-BESSE, UNIT 1 3/4 1-41 Amendment flo. M,61

This page left intentionally blank.

DAVIS-BESSE, UNIT 1 3/4 1-42 Amendment No. H. 61

O

~

This page left intentionally blank.

i l

l DAVIS-BESSE, UNIT 1 3/4 1-43 Amendment No, 6, 61 l

o I l3/4.2.

POWER DISTRIBUTION LIMITS l

l AXIAL POWER IMBALANCE I LIMITING CONDITION FOR OPERATION

'3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on l Figures 3.2-la,3.2-1b,3.2-Ic,3.2-2a,3.2-2 band 3.2-2c.

l APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.*

ACTION:

liWith AXIAL POWER IMBALANCE exceeding the limits specified above, either:

11

'a.

Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or b.

Within one hour reduce power until imbalance limits are met or to 40% of RATED THERMAL POWER or less.

iSURVEILLANCERE00(RL' PITS 16 ili;4.2.1.

The AXIAL POWER IMBALANCE shall be determined to be within limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% of RATED THERMAL POWER exceot when j,tne AXIAL POWER IMBALANCE alarm is inoperable, then calculate the AXIAL POWER i; IMBALANCE at least once per hour.

I i

i Ili 1

!!li ll

!i

!i

'i

-See so/5'iT Tes: Excection 3.10.1.

4 DAVIS-BE5SE, UN*T 1 3/4 2-1

-knendaent No. 33, /2, AE, 61 w

.~_ _ -

.c

Figure 3.2-la.

Axial Power Imbalance Limits. O to 24+10 -0 EFPD, Four RC Pumps - Davis-Besse 1, Cycie 4

(-17,iO2)

(20,iO2)

- 100

(-22,92)

(

- 90

(-25.80) 2--

80 (30,80) 5 m.

f-- 70 i5 60 PERMISSI BLE e

OPERATING q

RESTRICTED REGION REGION 50 T

s 8g--e m.

L.

l--

30 m.

20 IO I

I e

I

-30

-20

-10 0

10 20 30 Axial Power imoalance (7.)

DAVIS-BESSE, UNIT 1 3/4 2-2 Amendment No. JJ, A3,

/ 3, 61

t

\\

Figure 3.2-lb.

Axf al Power Imtalance Limii.s, 24 + 10. -0 to 150 + 10 EFPD, Four RC Pumps - Davis-Besse 1 Cycle 4

( 20,iO2)

(-21,iO2)

- 100

(-26,92)

(25.92) gg

(-29,80) g 80 (30.50)

W2 70 a

E PERMISSIBLE w

OPERATING A

60 RESTRICTED REGION S

REGION cr 50

.~

$E-- no

- 30 m

20 10 I

I I

I i

-30

. -20

-10 0

10 20

.30 Axial Power Imbalance (%)

DAVIS-BESSE, UNIT 1 3/4 0-2a

,centent No. 77, 33, A2, AJ,61

l Figure 3.2-ic Axial Power Imbalance Limits After 150 :10 EFPD, Four RC Pumps - Davis-Sesse 1, Cycle 4

(-26,iO2)

(20.iO2)

- 100

(~3I'82)

(25,92) 90 30

( 30,30 )

(-38,80)

_5 E

c 70 aI RESTRICTED 60 REGION PERMISSISLE A

OPERATING S

REGION E

50 o

T

_=

g_

S.

g-- 30 2

20 IO I

I I

I i

.uo 30 20

-10 0

10 20 30 Axial Power imealance (f.)

DAVIS-DESSE, Ui:IT 1 0/4 2-2b Amendment No. f;', f),61

This page intentionally left blank.

DAVIS-BESSE, Ut1IT 1 3/4 2-2c Amendment flo. 45,61 4

This page intentionally left blank.

DAVIS-BESSE, UtlIT 1 3/4 2-2d Amendment flo. AE,61

Figure 3.2-2a.

Axial Power Imbalance L'imits, O to 24 + 10, -0 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4

}

100

(-12.75,77) 80 (l5.77) i 2

~

( i6.5.69.5)

(i8.75.69.5)

.a (iS.75.60.5) f.-- 50

( 22.5.60.5)

=-

Oe c=

bo 40 RESTRICTED

-a REGION PERMISSIBLE $-

OPERATING b

REGION u

i--

20 si I

e i

-20 iO O

iO 20 30 AxialPowerimbalance(',)

I k

CAVI5-0 ESSE, UNIT 1 3/4 2-3 Amendment No. 77, 33, /E,61

~igur 3.2-2b.

Axial Power Imbalance Limits, 24 + 10, -0 to 150 + 10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 l

i 100

( i5.75,77)

(-19.5.69.5)

(ig,75,53,5) 5

(-2i.75,60.5) 2 - 60 (22.5.60.5)

=

a wA Sq uo c=

w Q

li$

8

~

RESTRICTED REGION PE. ISSIBLE

- 20 H

OPERATI NG REGION t;

3 c.

i f

f l

t i

-30

-20 iO O

10 20 30 Axial Power imoalancs (".)

CAVI5-CZ3-52, UNIT 1 3p :_;2 A :::.* ment flo. 77, 33, 47, /E,61

.U

l l

l Figure 3.2-2c.

Axial Power LThalance Limits After 150 =10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 1

100 i

- 80 (i5,77)

(-is.5.n)

(-23.25,6s.5)

(i8 75 SS 5) 2

(-28. 5. 60. 5)

W-- 60 (22.5.60.5)

P.

a ke i:

g--

40 tE -

cs o

PERMISSI BLE y

~

OPERATING REGION

$~

I

~

n.

REGION 5

E c

l 1

8 f

I

-30

-20 iO O

iO 20 30 Axial Power imbalance (f.)

DAVIS-BESSE, UNIT 1 3/4 2-3b Amendment No. 42,,45, 61

4 This page left intentionally blank.

f i

t l

l DAVIS-BESSE, UNIT 1 3/4 2-3c Amendment No. M, 61 i

e

l l

This page left intentionally blank.

DAVIS-BESSE, UNIT 1 3/4 2.3d Amendment No. fE, 61

3/4.1 REACTIVITY CONTROL SYST985 BASES 3/4.1.1 30 RATION CONTROL 3/4.1.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with costulated accident conditions are controllable within acceptable limits, and 3) tne reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

During Modes 1 and 2 the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insartion limits.

l SHUTDOWN MARGIN recuirements vary througnout core life as a function of fuel depletion, RCS boron concentration and RCS T"Ma.The most l

restrictive condition occurs at EOL, with_T at no d ooerating temoerature. Tne SHUTDOWN MARGIN required U9 consistent with FSAR safety analysis assumptions.

3/4.1.1.2 BORON DIUJTION A minimum flow rate of at least 2800 GPM orovides adequate mixing, prevents stratifitation and ensures that reactivity enanges will be gradual through the Reactor Coolant System in the core during boron concentration reductions in the Reactor Coolant System.

A flow rate of at least 2300 GPM will circulate an eouivalent Reactor Coolant System volume of 12.110 cubic feet in approximately 30 minutes.

Tne reactivity cnange rate associated with bo-on concentration reduction will be within the capability for operator recognition and control.

3/4.1.1.3 MCDERATOR TEMPERARJRE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that tne assumptions used in the accident and transient j

analyses remain valid through each fuel cycle.

The surveillance recuire.

ment for measurement of the MTC each fuel cycle are adequate to confirm the MTC value since this coefficient enanges slowly due principally to the reduction in RCS baron concentration associated with fuel burnup.

The confirmation that the measured MTC value is within its limit provides assurance tnat the coefficient will be maintained within acceptable values nroughout each fuel cycle.

CAVIS-3 ESSE, UNIT 1 3 3/4 1-1

l ! REACTIVITY CONN 0L SYSTE'.S l,

ii

!3ASE3 I

i b

i l3/4.1.1.4 MINDfUM TEMPERATURE FOR CRITICALITY liThis specification ensures that the reac' tor vill not be =ade cri:ical vi:h the

'reae:or coolant systa= average te=pera:ure less than 523*?.

This li=1:ation

'i.s required :o ensure (1) the =oderator temperature coeffician is within its

analyzed temperature range, (2) :he protec:1ve ins:rumentazion is within 1:s inor=al operating range, (3) the pressurizer is capable of being in an OPERABLE j s:atus with a s:eam bubble, and (4) :he reac:or pressure vessel is above 1:s

,l=ini=um RT D; ta=perature.

3 3/4.1.2.

3 ORA ION SYSTD'.S The boron injec:1on system ensures that negative reac:ivity control is avail; able during each mode of facili:y operation. The co=ponents required to per-form this function include (1) borated water sources, (2) makeup or DER pumps, l(3) separate flow paths, (4) boric acid pu=ps, (5) associated heat : racing

'sys:e=s, and (6) an emergency power sapp 1'y from operable emergency busses.

With the RCS average te para:ure above 200*?, a minimum of two separate and

iredcudant boron injec: ion systems are provided to ensure single func:1onal capabili:y in the even: an assumed failure renders one of the systems inop-erable. Allowable out-of-service periods ensure : hat =inor co=ponen: repair

.or correc:ive action =ay be comple:ad vi:hout undue risk to overall facili:y l safety from injection system failures during the repair period.

I lThe boration capability of either system is sufficient to provide a SF.' DOWN lMARGINfromalloperatingconditionsof1.024k/kafterxenondecayandcool-jdown to 200*?.

The maxi =um beration capabili:y requirement occurs from full jpower equilibriu= xenon conditions and requires the equivalent of either 7373 jgallons of 8742 ppm borated vacer from the beric acid storage :anks or 52,726

' gallons of 1800 ppm borated water from the borated rater storage tank.

,ne requirements for a minimu= contained volu== of 482,778 gallons of borated l

'va:er in the berated water s:orage tank ensures the capability for berating the.RCS to the desired level.

The specified quan:1:y of borated va:er is con-l isisten: vi:h the ECCS requirements of Specifica: ion 3.5.41 therefore, the

larger volume of bora:ed water is specified.

l jWith :he RCS te=perature below 200*?, one injection system is accept..ble with-j out single failure consideration on the basis of the i

I l.

ll DAVIS-fESSI, UNIT 1 3 2/4 1-2 Amendment No. Jr,3F,X,42,61

.y-w.

7 n-n..

., ~...,

+, -

(3/4.2.

POWER DISTRI3UTION LIMITS BASES i

The specifications of this section provide assurance of fuel integrity during

~

Condition I (normal operation) and II (incidents of moderate frequency) evues (a) maintaining the minimum DNBR in the core 2 1.30 during normal opera-by:

tion and during short term transients, (b) maintaining the peak lineas power density s 18.4 kW/f t during normal operation, and (c) maintaining the peak power density less than the limits given in the bases to specification 2.1 In addition, the above criteria must be met in during short term transients.

order to meet the assumptions used for the loss-of-coolant accidents.

The power imbalance envelope defined in Figures 3.2-1 and 3.2-2 and the insertion limit curves, Figuras 3.1-2 and 3.1-3 are based on LOCA analyses

- linear heat rate ruch that the maximum clad C

which have defined the --

temperature vill not exceed the Final Acceptance Criteria of 2200*F following Operation outside of the power imbalance envelope alone does not con-a LOCA.

stitute a situation that would cause the Final Acceptance Criteria to be ex-The power imbalance envelope represents the bound-caeded should a LOCA occur.

ary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined by Figures 3.1-2 and 3.1-3 and if Additional conservatism is the steady-state limit QUADRANT POWER TILT-exists.

introduced by application of:

Nucisar uncertainty factors, a.

b.

Thermal calibration uncertainty.

Fuel densification effects.

c.

d.

Hot rod manufacturing tolerance factors.

Potential fuel r'od bow effects.

a.

The ACTION statements which permit limited variations from the basic require-ments are accompanied by additional restrictions which ensures that the orig-2 inni criteria are met.,

l The ' definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:

Nuclear heat flux bot channel factor, is defined as the maximum local fuel FS rod linear power density divided by the average fuel rod linear power den-sity, assuming n W ami-fuel pellet and rod dimensions.

(

DAVIS-BESSE, UNIT 1 B 3/4 2-1 Amendnent No. JP,.3&,45

'"'7'~~T"r

o POWER DISTRIBUTION LIMITS BASES FN Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio M

of the integral of linear power along the rod on which minimun OfBR occurs to the average rod power.

It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimum DieR at full power are met, provided:

Fg 1 2.93; FN g 1,71 power peaking is nec a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMSALANCE produced by tne power peaking.

It has been detennined that the atove hot channel factor lim-its will be met provided the following conditions are maintained.

1.

Control rods in a single group move together with no indivicual rod in-sertion differing by more than E6.5% (indicated position) from the group average height.

Regulatikg rod groups are sequenced with overlaoping groups as required 2.

in Specification 3.1.3.6.

3.

The regulating rod insertion limits of Specification 3.1.3.6 are main-tain ed.

4 AXIAL POWER IMBALANCE limits are maintained. The AXIAL DOWER IMBALANCE is a measure of tihe difference in power between the too and bottom halves of the core.

Calculations of core cverage axial peaking factors for many plants and measurements from operating plants under a variety of ooerat-ing conditions have been correlated with AXIAL POWER IMBALAfCE.

The cor-relation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained between the limits specified in Soecifica-tion 3.2.1.

l The design limit power peaking factors are the most restrictive calculated at

} full power for the range from all control rods fully withdrawn to minimum al-j lowable control rod insertion and are the core DFSR design basis.

Tnerefo re, for coeration at a fraction of RATED THERMAL POWER, the design limits are met. When using incore detectors to make power distribution maps to oeter-N.

eine FQ and F 2H ens a.

The measurement of total oeaking factor Fg

, snall te increasec by

'..l cercent to account for manufacturing tolerances and furtner increasec :y 7.5 percent to acccunt for measurement error.

DAVIS-CESSE, Ut!IT 1 3 3/4 2-2 Amendment f'o. 77, 61

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