Text

Proposed Tech Specs,Increasing Fq(Z) & F-delta-H Peaking Factors,Incorporating Positive Moderator Temp Coefficient, Increasing Refueling Water Storage Tank/Accumulator Boron Concentrations & Increasing Sodium Hydroxide Concentration
	ML20205N395
Person / Time
Site:	Callaway
Issue date:	10/25/1988
From:	; UNION ELECTRIC CO.
To:	;
Shared Package
ML20205N389	List: ML20205N385; ML20205N391; ML20205N395;
References
	NUDOCS 8811040127
	Download: ML20205N395 (40)
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.. /i s Attachment ULNRC-1850 g

CYCLE 4 TECHNICAL SPECIFICATION CHANGES Section 3.1.1.3a page 3/4 1 Insert'l Section 3.1.2.5b.2) page.3/4 1-11 Section 3.1.2.6b.2) page 3/4 1-12 Section 3.2.1 pages3/42-d,2(a).

Section 3.2.2 s pages 3/4 2-4, 5, 5, 7, 7(a), 7(b)

~

, Section 3.2.3 page 3/4 2-8 Section 3.5.1c .page 3/4,5-l'

. Section 3.5.5b page 3/4 5-10 Section'3.6'.2.2a page 3/4 6-14 Bases Section 2.1.1 .pages D 2-1, 2 l Bases Section 3/4.1.2 pages B 3/4 1-2, 3 i

i Bases Section 3/4.2.1 page B 3/4 2-1 Easos Section 3/4.2.2 and 3/4.2.3 page B 3/4 2-4 Inscrt 2 Bases Section 3/4.5.5 page B 3/4 5-2 1

I, l

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! 8811040127 881025 PDR ADOCK 05000403

- P PDC

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

REVISION 2 REACTIVITY CONTROL SYSTEMS .

MODERATORTEMPERAlbRECOEFFICIENT LIMITING CON 0! TION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

a. L::: ;;-itiv; th;n ^ ai/k/*F f r th: :11 :d; wit,hd..r:rn,

, , , _ r e a i s < _ . _ _ _ _ ___2 i:;i--in;

__,4,.,. ..... . . .. .... v.u. .e no u. . . . . .anoc a.

b. Less negative than -4.1 x 10 4 Ak/k/'f for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.'1.1.3a. - HODES 1 and 2"#.

Specification 3.1.1.3b. - HODES 1, 2, and 3#.

ACTION:

a. With the MTC more positive than the limit of Specification 3.1.1.3a.

above, operation in MODES 1 and 2 may proceed provided:

NHhin //e above linIN 1.

ControlrodwithdrawallimitsarCestablishedandm sufficient to rastore the MTC to &;; ;;;itiv; than ; ak/h.'"T-within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANSkf within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Tnese withdrawal limits shall be in' addition to the insertion ,

limits of Specification 3.1.3.6;

2. The control' rods are maintained Within the withdrawal limits established above until a suosequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and

3. A Special Report is' prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured HTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive HTC to within its limit for the all rods withdrawn condition.

b. With the HTC oore negative than the limit of Specification 3.1.1.3b.

abeve, be in HOT SHUT 00VN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K,gf greater than or equal to 1.

5ee Special Test Exception Specification 3.10.3.

.I CALLAWAY

UNIT 1 3/4 1-4 I

INSERT 1 Less positive than +5 pcm/ F for power levels up to 70% RATED THERMAL, POWER and a linear ramp from that point to O pcm/*F at 100% RATED THERMAL POWER for the all rods withdrawn, beginning of cycle life (BOL) condition; and 1

N l

REVISq

3 :

RlACllVlfY CONTROL SYSTEMS e- ,

BORAILD WAILH SOURCF - SHUTDOWN

\

LlHITING CONDI110N FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be

'Ol'F R A3L E :

.i . A Boric Acid Storage System with:

1) A minimum contained borated water volume of 2968 gallt.,s,

2) Between 7000 and 7700 ppm of boren, and

3) A minimum solution temperature of 65'F.

h. lhe ref ueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 55,416 gallons, a3So

2) A minimum boron conce'itration of Mppm, and

3) A minimum solution temperature of 37'F.

AlPt(CARill!Y: MOOLS 5 and 6.

ARO_N:

With no burated water source OPERABLE, suspend all operations involving CORE AllLHAl!ONS or positive reactivity changes, h '/,[lltANCE REQUIREMENTS 4.1. 2. 5 The ahnve required borated water source ' hall s be demonstrated OPERABLE:

a. At least uncea per 7 <f ays. by:

1) Vori t ying the boron v.vncentrat ion est the water.

/) Ve r i t y l nt; t.ho cont ainett bor.1ted water volume, and I) Veritying the Buric Acid Storage System solution temperature when it is the murev of borated water.

D. At least unce per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it i, the source of borated water and the nutside 4If terperature is 1."'. th.in 3/*F.

4 N i AW ' t L'41 1 ; 3 ,' 4 !* !;

~

REVIS!on 3 - .

RIACllVilY CON 1ROL SYS1tMS Il0RAll 0 Wall W SOURCES - OPERATING LIM 111NG CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2 and 3 and one of the fol-lowing borated water sources shall be OPERABLE as required by Specifica-tion 3.1.2.1 for H00E 4: ,

a. A Boric Acid Storage System with:

L) A minimum contained borated water volume of 17,658 gallons,

2) Between 7000 and 7700 ppm of boron, and

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 394,000 gallons, 2:sv w d 2500

2) Between 2000 :nd 2100Vp pm of boron, *

3) A minimum solution temperature of 37'F, and T

A maximum solution temperature of 100*F. '

4)

APPLICABil)1Y: M'0E5 1, 2, 3, and 4 ACTION:

, a. With the Boric Acid Storage Systen, inoperable and being used as one ni the above required borated water sources in MODE 1, 2, or 3, restore the storage system to OPERABLE status within 72 hourt or be in at least Il0i STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTOOWN MARGIN equivalent to at least l% ok/k at 200'F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHuiOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With the RWST inoperable in MODE 1, 2, or 3 restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANOBY within the next s hours and in COLD SHUTDOWN witnin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With no horated water source OPERABLE in MODE 4 restore one borated water source to CPERABLt status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLO 5HUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

CAttAWAY - UNil 1 3/4 1-12

AERoX TELZecPIEA 495 ; 3-11-o7; 13 :N : + 314075co4G e o H0V.03 '87 13:25 HRC ME5 SAGE CENTER BETHESDA MD P.003 3/4.2 POWER O!STRIBUT!0N LIMITS 3/4.2.1 AX!AL FLUX O!FFERENCE LIMIT!NG CON 0! TION FOR OPERAT!0N _

3.2.1 The indicatec AX!AL FL"X O!FFERE.'i E (AFO) shall be maintained wita.in the following target band (flux difference units) scout tne target flux difference:

a. +3t. -12t for Norr.41 Oparatf or

b. +3%

for RESTRICTED AFD OPEMTIO.*( ,

greater than or equal to 50t tut less than 0.9 AFL}aThe or 90%indicated of AATEDAFD T .EF/A cay deviate outside t POWEA, whichever is less, provided the indicated AFD is within tne Acceptacle i Operation Limits of Figure 3.21 and the canulative penalty deviation times dcas not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD may deviate outside the applicabla required target band at i greater than 155 but less thsti 505 of MTED THEML PO*='ER provided the cumula-

, tive penalty deviatien time coes not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

. APPLICA3!LI'TY: NOCE 1, above 15% of MTED THEWL 70WER*.i ,

ACTION:

a, 'dtth the indicated AFD outside of the applicable required tarcet ;

i band and with THERMAL POWER greater than or equal to 0.9 AFLND" or 90% of MTED THE;M. POWER, whichever is less, within 15 minutes, either:

1. Restore the indicated AFD to within the applicable required I target band limits, or

' 5ee Special Test Exception Specificaticn 3.10.2.

Surveillance testing of the Power Range Neutron Flux channel .ay be :er.

fomed pursuant to specification 4.3.1.1 provided the indicated AFD is reintained within tba Acceptable Operation Limits of Figure 3.2-1 ano r THERMAL POWER dPL uhm . A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operstion may be ac:amulatec with the AFD outside of the acolicable required target band during testing without penalty deviation.

" APLNO is the minimus allowable power level for RESTRICTED AFD OPEMTION and will be provided in the Peaking factor Limit Recort per Specification 6.9.1.9.

apt.NO is e:ual to the a,So

.inimum 'M' V U2) ,[

over Z Lfj(2)* e'(2 )so l and Fj(2) and W:)NO are defined in a.2.2.2.:.

CALLAMAT - UNIT 1 3/4 2 1 A wncment No. 28

POWER O!STRIBUTION LIMITS .

SURVEILLANCE REQUIREMENTS 4.2.1.2 The indicated AFD shall be considered outside of its target band when two or corn GPEFJELE ex::re channels are indicating the AFD to be outside the target band. Fenalty deviation outside of the above required target band .

shall be accumulated on a time basis of:

a. One =inute penalty deviation for each 1 minute of POWER OPEFATION outside of the target band at THE??AL POWEP. levels ecual to or acove E0'. of FA!ED THEF.s.L POWER, and

b. Or.e-half cinu:e penalty deviation for each 1 minute of 70WER OFE?A-T10*i outsice of the target band at THEFFAL PCWER levels between 155 and 50L of FATED THER"AL PCWER.

4.2.1.3 The target flux difference of each OPERA 5LE excere channel sna11 Tr.eee de: ermined oy ceasure ent at less; on:e per 92 Effective Full Pcwer Days.

cr: visions of Specification 4.0..'. are not a::licable.

4.2.1.4 The target flux difference shall be updated at least once :er 31 Effec-tive Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or at by thelinear interpolation end of the cycle between life. the mostThe provisions recently measured value and of Specification 4.0.4 are not applicable..

+he calcu/Jad wlac A

3/4 2-2(a) A ene ent No. 23 CALLAWM - UNIT 1

. l

REVISlay 3-POWLW 0151R100l!0N LIM 11$

3/4.2.2 llCAITLtlX1101 CHANNEL. FACTOR-Fg g LIMITING CON 01110N FOR dPERATION 3.2.2 F (Z) shall be litited by the following relationships:

9 J.ro FA (Z) ~< [ T ] (K C )] for P > 0.5, and Fg (Z) 1 ( . ] (K(Z)] for P 1 0.5._.

Where: S.40 p , THERMAL POWER , and RATED THERMAL POWER K(1) = the function obtained from Figure 3.2-2 for a given core height location.

APPLICABillTY: MODE 1.

AC110N:

With Fq (Z) exceeding its limit:

a. Reduce THERMAL POWER at least M for each M9 F (Z) exceeds the 1mit within 15 minutes and similarly reduce the Power Range Neutron Flux-

'}

High Trip Setpoints within the next 4 bt.urs; POWER OPERATION may proceed fce up to a tota'i of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; tubsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 3 for each n F (Z) q exceeds the limit; and

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL. POWER may then be increased provided Og (Z) is demonstrated through incore mapping to be within its limit.

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l CALLAWAY - UNIT 1 3/4 2-5 I

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4 P] aid. DIST4*BUT!0'i LIMITS SL'D.'!EILLfNCE REOUIREMENTS , _

~

4.2.2.1 The provisions of S ecification 4.0.4 are not applicable. ,

4.2.2.2 f:r turnal C:erati:r., Tg':) tr.zil te e.aluated t: dete .ine if Tg(z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution cap at any THIP"AL P2WED greater than 5% of PATED THEP?AL POWER.

1

b. In:reasing the ceasured FQ(:) cer;:nent of the power distrioution

- rao by 3i :c ac::unt fcr ranufs:turin; toleran:es and further in:reasing tne value ey 5!, to acc unt for ceasure .ent un:ertair. ties,

c. Satisfyir.; the B'.1: sin; relati:ns:ii::

2.50 Fg (z) 1MFx d :) f c- P > 0. 5 M

F x n.:;:.:

.2.50 Tg (:) < G--Mx Q f o r P 1 0.5 M

W(z)f;) x 0 f gl2.fo where fg (z) is the measured Fg(z) increased by tn; allowarces for manufacturing)

Tolimit,X(2 tolerances and ceasurement uncertainty. G-OS POWEP, and W(z)NO is the cycle dependent, tiomal Operation furetion tnat accounts for power distribution transients encountered during Normal Operation. Tnis function is given in the Feu.f'1g factor Limit Report as per Specification 6.9.1.9.

l

d. l'.easuring FgM (z) according to the following schedule:

l. Upon achieving equilibrium conditions after exceeding, by 10'4 or more of FATED THERFAL POWER, the THERPAL POWER at which FQ (z) was last detemined,' or

1. At least once per 31 Effsetive Fu,ll Power Days (EFPD), whichever l occurs first.

^

During po er escalation at the be; inning of each cycle, power lasel eay be increased until a power level for extended operatir: '" heen achieved and a power distribution cap obtained.

3/4 2-6 /c.!ndment tio. 28 CALLAWAY - UNIT 1

. 5 i

P02ER 015TRIBUTION LIMITS SURVEILLL' ICE REQUIREMENTS (Continued) 4.2.2.2 (Continued)

e. With maasurements indicating maxieum f5(z) over : Ktz) ;

, M nas in:reased sin:e tne previ:us determination of FQ(z)',either of :ne following a:,ticas sns11 be taxen: I

1. FJ9:) shall be increased by 2t over that specified in ;

Sie:ification 4.2.2.2:., er t

2. F[(z) shall be measured at least once per 7 Effective Full F:,4er Days until two successive mass indicate tha: l i

is not increasing.

maximum Th(z) .

over z y

f. With the relationsnips specified in Specification 4.2.2.2c. above

' not being satisfied: -

2

1. Calcule,te the percentQF (z) exceeds its limit by the follcwing expression:

) -1 x 100 for P 1 0.5 (max. over z of F$(z) x W(z)NO M.'s :

x 100 for P < 0.5 l

(max. over z of F((z) x W(z)NO )-1 7 X(z)

_ dx 2.S0 i

2. Either one of the following actions shall be taken:

[

(a) Comply with the requirements of Specification 3.2.2 for

]

Fn(z) exceeding its limit by the percent calculated noeve, or !

(b) Verify that the requiremen'.s of Specification 4.2.2.3 [

. for RE:TRICTED AFD OPERATION are satisfied and enter RESTRICTED AFD OPERATION.

and ,

g. The limits soecified in Specification: 4.2. 2. 2. c . , 4. 2. 2.2. e. i j

4.2.2.2.f. at.ove are not applicable.'- ie following core plane l regions:

Lover cora region from 0 to 15; inclusive.

i 1.

2. U;;er core region from 35 to 100t. inclusive.

I A, .enement No. 23 sat.LAAY - UNIT 1 3/4 2 ~

t I

_ - _ _ _ _ _ , _ . _ _ . - _ . , _ _ _ _ . _ . _ _ _ . _ _ _ _ _,,--_m_ _ _ _ . , _ _ _ , _ _ _ _ . . _, - _ - , __ _ _ .

POWER DISTRIBUTION LIMIT,5_

SURV!ILLANCE REOUIREMENT5 (Continued)

ND 4.2.2.3 RESTRICTED AFD OPEPATION (PAF00) is permitted at powers above APL if the following conditions are satisfidd:

a. Prior to entering RAFDO maintain THEFFAL PCWE'R above APLND and h ss tnan or equal to that allo ed by 5;ecification 4.2.2.2 for at le at tne crevious 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain FAFCC surveillance (AFD within FAF00 +2:

is tnir.

of target flux differen:e) during this time ;eriod.

per:-i:: d :reviding TxEP.'ML POWER is maintained bstween AFLND and APL "' W or between APLO and 100t (wnienever is more limiting) and F[l.re A.-

JrgyPiliance is oefinec as: is caintainsi :ursuar.: to Seedficatien 4.2.2.4 3 50 ~

APL""" = minimum ~G-#x Kf:) x 1001 ver : ,,7Q M

( ) x g(g) where: Fy(z) is the censured Fn(z) increased by the allcwances for The FQ limit d'go' .nufacturing)tolerancesandmeasurementuncertainty.

is . X(z is given in Figure 3.2-2. W(7)PAF00 is the cycle dependent function that accounts for limited power distribution transients encountered during RAF00. This function is given in the Peaking Factor Limit Feport as per Specification 6.9.1.9.

b. During RAFDO, if the THETWAL POWER is decreased below APLND then the conditions of 4.2.2.3.a shall be satisfied before re-entering PAF00.

4.2.2.4 During RAF00. Fg(z) shall be evaluated to deter nine if Fg(z) is within its limits by:

a. Using the movable incore detectors to obtain a power dir.tribution map at any THEF#AL POWER above APLND.

b. Increasing the measured F 0 (z) component of the power distribution map by 3% to account for miufacturing tolerances ano further increasing the value by 5% to account for measurement uncertainties,

c. Satisfying the following , relationship:

.?. 50 NO rM Vx Kf:) for P > APL

] 1 +P x W(z)RAFM F'(:) 2.SO.

where: F ) is the .vasured F (:). The Fo limit is +r46.VX (:)

P is Qthe relative THEFFAL F0WER. W(:Jpjpg isgiven$(: in Figure 3.2-2.

is the cycle decendent function :ns: secounts for ifmiteo power distribution transients encountered during RAFDO. This function is given in the Peaking Factor Limit Report as per Specification i.9.1.9.

2 N 2.J(a) Amenomen; No. 23 CALLAWAY - UNIT 1

POWER 015TRIBUT!0N L!t41TS SURVEILLA!;0E REOUIREMENTS (Continued) 4.2.2.4 (Continued)

d. P,tasuring F$(2) in conjunction with target flux ('lfference detemi-nation according to the following schedule:

1. Prior to entering RAF00 after satisfying Section 4.2.2.3 unless a full c:re flux mac has been taken in the previous 31 IFPD wip the relative inertal power having been maintained at:ye APL~, f:r tr.e 24 n:urs prior to caS;fng, an:

2. At least on:e per 31 Effective Full Po ar Days.

e. With measure: ants indicatin; riximun Fg(:) ,

g M:) ; ,

v has increased since the previous determinatien of Fy(:) either of the folicwing actions shall be taken:

1. Th(z) shall be incr;ased by 2 percent over that specified in 4.2.2.4.c, or

2. Fh(z) shall be measured at least once per 7 EFPD until two successive maps indicate that maximum ~h(z)'isnotincreasing.

F nr: g(;)

f. With the relationship specified in 4.2.2.4.c above not being satisfied, comoly with the requirements of Specification 3.2.2' for F Q(z) exceeding its limit by the percent calculated with the following expression: _

ND (max. over z of Fh(:) x W(:)RAFD0 ) -1 x 100 for P t APL

.:' x X(:)

?

L 2.So -

9 The limits scecified in 4.2.2.4.c. 4.2.2.4.e, and 4.2.2.4.f above are not apolicsole in the following core plane regions:

1. Lower core region from 0 to 15 percent, inclusive.

2. Upper core regien from 35 to 100 percent, inclusive.

4.2.2.5 When Fn(:) is measured for reasons other than meeting tne recuire-ments of Specification 4.2.2.2 or 4.2.2.4, an overall easured F Q (:) snall be cotained from a cower distribution mac and increased oy 2t to account for

.anufacturing tolerances and further increased by 5t to account for measure-ment uncertainty.

3/4 2-7(b) A enc ent No. 23 CALLAWAY - U.'ilT 1

POWER OfSTRTBUTTON LfMfTS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F g LIMITING CONDITION FOR OPERATION

3.2.3 Fhshallbelimitedbythefollowingrelationship

/.59 Fh 1 .

[1 + 0.3 (1-P))

~~

where P = THERMAL POWER RATED THERMAL POWER Fh=MeasuredvaluesofFfH obtained by using the movable incere detectors to obtain a power distribution map. The measured values of F g shall be used since an uncertainty of 4% fo*

incoremeasurementofFhnasbeenincludedintheabovelimit.

APPLICABILITY: MODE 1 ACTION:

With Fh exceeding its limit:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. RestoretheFhtowithintheabovelimits,or

2. Reduce THERMAL POWER TO LESS THAN 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to i 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Demonstratethroughin-corefluxmappingthatFhiswithin its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATGD THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may pro.

ceedprovidedthatFhisdemonstratedthroughin-coreflux ,

mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL l POWER.

l CALLAWAY - UNIT 1 3/4 2-8 Amencment No. 15

i 3/4.5 Ew!RG!hcY CORE COOLING SYSTEMS 3/4.5.1 ACCUMJtA1CRS LlHITING CONDIT1CN FOR OPERATION 3.5.1 Eacn Reactor Cc:Isnt Syste- :cu uiator shall be OPERASLE with:

a. The isolatter valve open and pc-er removed, '

l O. A ccatainec horated =atar volume of betmeen (061 and 6555 ss11ons.

l

c. A boron ecncentratien of between Mirirgand 4M&g p;m, ar,d

.3.1M 2 S60

c. A nitrogen cever :ressure of etween 602 and 643 psig.

A: 0.l CL5 : LIT): M30E5 1, 2. an: 38 ACTIO,:

5 a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within I hour or be in at least NOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in NOT SHUT 00hH within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b, kitn one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDSY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTD0hw within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l 2

SURVElttANCE REQUIREMENTS a

I 1 4.5.1.1 Each accumulator shall be demonstrated OPERASLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and

2) Verifying that each accumulator isolation valve is open, "Pressurizer pressure above 1000 psig.

Ar,end ent No. 28 3/4 5-1 CALL AVAY

UNIT 1 4

l REVISfoy y ~.

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EMERGENCY CORE COOLING SYSTEMS

'N ;

l 3/4.5.5 REFUELING VATER STORAGE TANK i

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LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with: ;

a. A minimum contained borated water volume of 394,000 gallons, 2350 *nd afoo b.* A boron concentration of betw' ten 0000 . 4 0100Mvpm of boron,

c. A minimum solution temperasure of 37'F, and

d. A maximum solution temperature of 100*F.

APPLICABILITY: H00E5 1, 2, 3, and 4 ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the -

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,

~.

5 SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the contained borated water volume in the tank, and

2) Verifying the boron corcentration of the water,

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air terparature is either less than 37*F or greater than 100*F.

J CALLAWAY - UNIT 1 3/4 5-10

CONT AINMENT SYSTEMS . S/Qg y SPRAY ADDITIVE SYSTEM .

i LlHITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with:

a. A spray ' additive tank containing a volume of between 4340 and 4540 gallons of between 23 = i 31%gby weight NaOH solution, and gg

b. Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a Containment Spray System pump flow.

APPLICABILITY: H0 DES 1, 2, 3, and 4. -

ACTION:

With the Spray Additive System inoperable, restora the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the Spray Additive System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE: ,

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

b. At least once per 6 months by:

1) Verifying the contained solution volume in the tank, and

2) Verifying the concentration of the NaOH solution by chemical analysis.

c. At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-High-3 (CSAS) test signal; and

d. At least once per 5 years by verifying ,

1) Each eductor flow rate is greater than or equal to 52 gpm using the RWST as the test source throttled to 17 psig at the eductor inlet, and

2) The lines between the spray additive tank and the eductors are not blocked by verifying flow.

CALLAWAY - UNIT 1 3/4 6-14 ,)

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2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the r,elease of fissi products to tne reactor coolant.

4 by restricting fuel c;aratien to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface te-:arature is sli;htly a :ve :he c: lant satura:i:n tem: stature.

' 0:eration above the u;;er beundary of tne nuclease boiling regime ceuld result in ex:essive claddin; tem:eratures because of tne en C';3 is not a directly censurable parameter durin; cperation and cesfficien .

tnerefore THEFy.*L POWER and Reactor Cc:lant Te perature and rressure have be

This relation has been developed to predict the ONS flux and 4 related to 0N1.

r The local DNS heat flux ratio (ONSR) definedi as the ra would cause DNS at a particular care location to the local heat flux, is ind c-ative of the margin to ONS.

> there must be at least a 95 percent The DNS design basis is as follows:

l probability that the minimum DNSR of the limiting rod during Condition I events is greater than or equal to the DN3R limit of the OSS correlation being used (the WRS-1 correlation for Optimized fuel (OFA) and the WRB-2 correlatio for VANTAGE 5 fuel in this application). The correlation DNBR limit is estab-lished based on the entire appifcable experimental data set such that there is when the minimum 043R is at the ONBR limit (1,17 WRS-2 for both cor'r elations) .

nuclear and thertnal parameters, and fuel fabrication par statistically such that there is at least a 95% probability with 95t confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNSR limit. The uncertainties in the above plant parameters are used to detertnine the plant DNBR uncertainty.

correlation DNBR limit, establishes a design DN3R 2 d ' S for th- mble and typical cells.

For Callaway, the design 0*iBR value's are ' . :

cells, resoectively, for OFA, an@In addition, margin has been maintained in

- respectively, for VANTAGE 5 fuel. :yanalysisONBRlimihsof1.42and1.45for

' both fuel designs by meeting safe l thimble and typical cells, respeg:ively, f r AN E for uel. 0FA, and(n.51 g g g j,y and 1.59 for th

and typical cells, respectively, The curves of Figure 2.1-1 shew the loci of coints of THE?#.tl POWER, j Reactor Coolant System pressure and average tem: eratur liquid.

st the vessel exit is less inan :ne en:nalpy of saturate i

l A enc ent No. 75, 23 3 ;-1 c;4LAWAy . unit 1 I

I 1

SAFETY LIMITS ,

8ASES

(~ for- 0FA ed /.59 Ar VAWrAGE S fue/>

2.1.1 8 EACTOR CORE (Continued) , .

measured The curvesarebasedonaVnuclearenthalpyrisehotchannelfactor,Fh, j 2 of 1.4904nd a reference cosine with a peak of 1.55 for axial power shape. An i allowance is included for. an increase in Fh at reduced power baseo on the j expressions; [

t F = 1.49 (1+ 0.3 (1-P)] for OFAj sn/

' wh'ere P is the fraction of RATED THERMAL POWER. ;

These limiting heat flex condi,tions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control :

rod insertion assuming the axial power imbalance is within the Itaits of the f (AI) function of the Overtemperature trip. When the axial power imbalance ,

is not within the tolerance, the axial power.f ebalance effect on the Overten- .

l . perature AT trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE j The restriction of this safety limit protects the integrity of the Reactor l Coolant System (RCS) from overpressurization and thereby prevents the release :

of radionuclides contained in the reactor coolant from reaching the containment l atmosphere, ,

The reactor vessel, pressurizer, and- the RCS piping and valves are designed to Section III of the ASME Code for Nuclear Power Plants which i i

permits a maximum transient pressure of 110% (2735 psig) of design pressure.

i The Safety Limit of 2735 psig is therefore consistant with the design criteria and associated Code requirements. ,

The entire RCS is hydrotested at' greater than or equal to 125% (3110 psig) of design pressure to demonstrate integrity prior to initial operation.

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l 3 2-2 .wenesen: No. 15 f CALLAWAY - UNIT 1 i

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, y REVislog I a 1EA,C1!VITY CONTROL $YSTEMS .

3

.- r.-- -

MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MOC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.1 x 10 4 ok/k/**. The MTC value of -3.2 x 10

ak/k/'F represents a conservative value (with correc-tions for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10 4 ok/ k/* F.

The Surveillance Requirements for measurement of the MTC at the beginning

.ind near the end of the fuel cycle are adequate to confirm that the HTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

1/4.t.1.4 MINIMUM TEMPERATURE FOR CRITICALITY Ihis specification ensures that the reactor will not be made critical -

with the Reactor C60 tant System average temperature less than 551'F. This limitation is required tu ensure: (1) the moderator temperature coefficient is within its analyzed temperature range (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT NOT temperature.

1/4.1.2 BORATICN SYSTEMS Ibe Boration Systems ensure that negative reactivity control is available iluring each MODE of facility operation. The components required to perform this lunctinn include: (1) borated water sources, (2) centrifugal charging pumps, ,

( 't) ' epar.it e flow paths (4) boric acid transfer pumps, and (5) an emergency power .upply from OPERABLE diesel generators.

With t he RCS everage temperature equal to or greater than 350'F, a minimum of two bnron injection flow paths are required to ensure single functional t.ipab i l i t y in the event an assumed failure renders one of the flow paths i ntipe r.4h l e , the horation capability of either flow path is sufficient to provide a MittlDOWN MARGIN f rom expected operating conditions of 1.3% Ak/k af ter xenon elecay and coohtown to 200'F . The maximum expected boration capability requirement occurs at LOL f rom full po.er equilibrium xenon conditions and requires

11.653 4miluns of 7000 ppm borated water from the boric acid storage tanks cr ~

83./45 qallons of .eM& ppm borated water from the RWST. With the RCS average temperature less han 350*F, only one boron in.4ection flow path is required. , ,

23 % ..J CAllAWAY - UNIT 1 S 3/4 1-2

PEACTIVITY CONT"40L SYSTE% 4 i BASIS 50 RATION SYSTE!!S_ (Continued)

With the RCS temperature below 200'F one Beration' System is acceptable ;

without single failure consideration on the basis of the stable reactivity '

condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron 1 Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pu p to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 provides assurance that a mass addition pressure transient can be relieved by the operation of a sing'le PORV or an RHR suction relief valve.

4 t

The boron capability required below 200'F is sufficient to provide a SHUTDOWN MRGIN of 1% ok/k af ter xenon decay and cooldown from 200'F to 1409F. ,

This condition requires either 2968 gallons of 7000 ppm borated water from the ;

boric acid storage tanks or 14.076 gallons of 300gppm borated water from the RWST.

235o l I

The contained water volume limits ir.clude allowance for water not availabl because of discharge line location and other physical characteristics.

The limits on contained water volume and boren con:entration of the RWST l also ensure a pH value of tetween 8.5 and 11.0 for the solution recirculated This pH band minimizes the evolution of

. within Containment af ter a LOCA.

iodine and minimizes the effect of chloride and caustic stress corrosion on l mechanical systems and components. F i

The OPERABILITY of one Boration System during REFUELING ensures that this i system is available for reactivity control while in PODE 6.

3 /4.1. 3 M3VABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power

! distribution limits are maintained. (2) the minimum SHUTDOWN MRGIN is m ;

tained, and (3) the potential effects of rod misalignment on associated acci- "

l dent analyses are limited. OPERABILITY of the control rod position indicators is required to detemine control red positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within + 12 steps at 24, 48,120 and 228 steps withdrawn for the Control Banks and li, j

' 210 and 228 steps withdrawn for the Shutdown Banks provides assurance that the ;

Digital Rod Position Indicator is operating correctly over the full range of i indication. Since the Digital Rod Position System does not indicate the actual 1

i shutdown rod positien between 13 steps and 210 steps, only points in the indi- t cated ranges are picxed for verificatim af agreement with to :aded position. l

! Shutdown and control rocs are positioned at 225 steps or higner for fully witn- ,

J drawn. :

3 3/4 l-3 A"4Dd ent No. 29 ;

CALLAWAY - UNIT I i

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. : ... .s. .. . . . . - . . . . :~,,_..._._...%.. -

........:.. - .:.. , . .. :'.L . . ~ v - - '

p0WER O!STRIBUTION LIMITS 3/4.2 _

BASES The specifications of this section providef Moderate assurance Frequehey of fuel integrity

=during condition I (Non.al Operation) and II (Incidents o events by:

-(1) maintaining the minimum DNSR in the core at ord above the analysis DNBR limits during nomal operation and in short-term trans (2) liniting the fission gas release,i fuel i pellet In addition, temperature, limiting and c techanical pecperties to within assumed design cr ter a. ides assurance that the peak linair pcaer density curing Ccnditien I events prov d the ECCS tne initial ccnditions assumed for the 1.0CA analyses are cet an acceptance criteria limit of 2200*F is not exceeded.

The definition of certain h'st channel and peaking factors as used i these specifications are as follows:

F(Z) g Heat Flux Hot Channel Factor, is defined as fuel rod heat flux, allowing for manufacturirg tolerances on fuel , , .. .

pellets and rods; and '

d I aH Nuclear Enthalpy Rise Hot Channel Factor, is !

power to the average rod power.

3/4.2.1 AXIAL FLUX O!FFERENCE_ d The limits on AXIAL FLUX OlFFERENCE (AFD) d d during assure that the i

e,nvelopesof and .2.So 2.324 fe,. ofAtimes Q gqy-Mgg the normalized S) r.aye c}rve/p axial changes. The Target flux difference is determined at equilibrium xenon ition conditi full-length rods my be positioned within the core in accordance'w respective insertion limits and shouldThe be value inserted of thenear targettheir flux normal for steady-state operation at high power levels.

difference obtained under these conditions divided by the fraction THEp."AL POWER is the target Target flux fluxdifferences difference for other at RATED THER."ALTHERMA associated core burnup conditions. POWER levels are obtained appropriate f ractional THER."AL POWER level . flux differen The limits on AXIAL FLUXOne O!FFEFiENCE mode is Nomal Operation, (AFD) where arethegiven in Two modes of operation are pemissible. The AFD limit for this applicable AFO limit is defined by Specification 3.2.1.a. flux difference. i mode of operation is a +3,

-12t target band about the targetAfter extended FQ (Z) less thanload fo tions in the maximum allowed power to guarantee operaticn w its limiting value.

Amendment No.15, 28 8 3/4 2-1

,CALLAWAY - UNIT 1

PO*1ER DISTRIBUTION LiHITS BASES 3/a.2.2 and 3/4.2.3 LT / LUX HOT CHANNEL FACTOR AND NUCLEAR ENTFALFY RISE fiOT CHAfi.'iEL FACTOR (Continued)

- Each of these is censurable but will n:r:sily onl ically as spe:ified in S;ccifications 4.2.2 and 4.2.3.* Thisyperiodic.

be detemined serve 11- period-lance is sufficient to' ensure that the limits are taintained provided:

a. C:ntrol r:ds in a sic;1e gr:vp m:ve together with no 'individal

< rod insertion differing by c:re than :12 steps, indicated, fr:m tne group de .and position.

2. C: .:r:1 red banks are se:;;en:ed witn everla;;in; gr:v;s as des:ribed in Specification 3.1.3.6.

c. The control red insertion limits of See:ification 3.1.3.6 are maintained,

e. ..

d. The axial power dis'tribution, express'ed in tems of AXIAL FLUX DIFFERENCE, is maintained within the limits. ..

N aboveafge will be maintained within its ligits provided conditions a. through d.

F maintained. The relaxation of F allcws changes in the radial power shape fd, as a function of THERMAL POWER When an Fn measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5: is appropriate for a full-core map taken with the incere detector flux capping system and a 3:

allowance is appropriate for e.anufacturing tolerance.

N is measured, (i.e., inferred), no additional allowances are When F necessarypr$ gor to comparison with the Ifmits of Section 3.2.3. . An error allow-ance of 4: has been included in the limits of Section 3.2.3. .. -.,, ;. g g A 3.T ad4 M m .

.7 between the safety analysis DNSR limits -(1.42'and 1.45 for the ,g g ZNJ'ER*T' 2 ----+in Marg

Optimized fuel thimble and typical cells, respectively, ard 1.61 an01.69 for -

the VANTAGE 5 thimble and typical cells) and the design D! BR limitst"/ . : e d

+r3+ for the Optimized fuel thimble and typical cells and%;; ;;d 1.33 for the VANTAGE 5 thimble and typical cells, respectively) is maintained. A fraction i

a of this margin is utilized to acconnodate the transition core DNBR penalty

- (2' f:r 0;tici::d '-412ht for VANTAGE 5 fuel) and the appropriate fuel red bow Dh4R penalty (less than 1.5t per WCAP-8691, Rev.1). The margin betwean 77,g*

design and safety analysis triGR limits of for Optimized fuel and .or VANTAGE 5 fuel includes greater than 3: rar n for Optimized fuel and 49-

="i- Sr VANTAGE 5 fuel for plant design d.3%lexibiity.bedh y

The hot channel factor F'(z) Q is neasured periodically and increased by a cycle and height dependent pe=er factor appropriate to either Norval Operation to provide assurance that the or limitRESTRICTED on the hot channel AFDfactor, OPEPATION, z) is r.e.. W(z)k(or W(z)pp,W(z)33 acceunts for the e CALLAWAY - UNIT 1 B 3/4 2-4 kencrent No. Lf. 28 S

e.*p ,We -ge e, .g g e ,. 9%* e sea e9** *** pq g g aqm g,. g

** N *f. egg *w ea ga, se e e e g e e e gn omewque e g, , ,, ee, 9ge

t i

I l

INSERT 2 Specifications 3.2.2 and 3.2.3 contain the F g and F-delta-H limits applicable to VANTAGE 5 fuel. The OFA fuel is l analyzed to lower limits since it will have experienced l burnup, thereby reducing the attainable OFA-specific hot

channel factors such that the expected peak power levels and peak radial power of the OFA fuel will be much less than that necessary to approach the OFA Fg and F-delta-H analysis limits.

l l

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Lm m.m. <0.n.Cwun.w.n M, REVISIcy j .

'.W. =:-- - = =-

ECCS 'UB5YS1 EMS (Continued)

The limitation for a maximum of one centrifugal charging pump to be

< OPERABLE and the Survet)1ance Requirement to verify all charging pumps and i d

Safety Injection pumps except the required OPERABLE charging pump to be inoperable in M00L5 4 and 5 and in MODE 6 with the reactor vessel head on provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or RHR suction relief valve. -

The Surveillance Requirements provided to ensure OPERABILITY of each component ensure, that at a minimum, the assumptions .nsed in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements i

for throttle valve position stops and flow balance testing provide assurance a that proper ECCS flows will be maintained in the event of a LOCA. Maintenance l , of proper flow resistance and pressure drop i,n the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding J runout conditions when the system is in its minimum resistance configuration,

(?) prnvide the pruper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level uf tutal ECCS flow to all injection points equal to or above that assumed in the LCCS-LOCA analyses. The Surveillance Requirements for leakage testing of [CC$ check valves ensure that a failure of one valve will not cause an intersystem LOCA. The Surveillance Requirement to vent the ECC5 pump casings and accessible, i.e., can be reached without personnel hazard or high radiation .h dose, discharge piping ensures against inoperapie pumps caused by gas b,inding '

/

or water hammer in ECCS piping.

}/4.5.5 REFUELING WATER STORAGE TA4K The OPERABILITY of the refueling water storage tank (RW57) as part of the LCCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The 1tmits on RwST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor wil'l remain subcritical in the cold condition following "---' '""--^

mixing of the RWST--'and the-*--' RCS water M use s " ' '

m B ! q 1hese assumptions are consistent with the LOCA analyses.

, n ruonim s// M e con +n/ edt are ou+ of Me cor e .

The contatiTed water volume limit includes an allowance for water not uuhle because of tant discharge line location or other physical characteristics.

the limits nn cuntained water volume and boron concentration of the RWST al'.o ensure a pH value vf between 8.5 and 11.0 for the solution recirculated within containment atter a LOCA. This pH band minimizes the evolution of iosline .uwt minimites the of fect of chloride and caustic stress corrosion on wthanical systems and components, CALLAWAY - UNif 1 B 3/4 5-2 e e4i

9 e-

Attcchment 5 ULNRC- 1850 SIGNIFICANT HAZARDS EVALUATION FOR CYCLE 4 l

i I

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4 tW SIGNIFICANT HAZARDS EVALUATION FOR '

CALLAWAY CYCLE-4 RELOAD l

I. INTRODUCTION This ovaluation supports Union Electric Company's license amendment request for the Callaway Plant Cycle 4 Reload.

Changes in Cycle 4 which require revision of the Callaway Technical Specifications include increased peaking factors, incorporation of a positive moderator temperature coefficient (PMTC), increased refueling water storage tank (RWST)/ accumulator boron concentrations, and increased spray additive tank (SAT) sodium hydroxide concentration.

Sections II, III and IV of this evaluation discuss cach of l these changes and provido an analysis using,the standards of 10CFR50.92 (i.e., the three factor test).

II. INCREASED PEAKING FACTORS In Cycle 4 Union Electric proposes to increase core peaking factors to allow more flexibility in the fuel management schemes by allowing longer fuel cycles and improvement of I fuel economy and neutron utilization. The limit on the measured nuclear enthalpy risc hot channel factor (F-delta-H) will increase from 1.49 to 1.59 and the limit on i the heat flux hot channel factor (F (z)) will increase from ;

2.32 to 2.50. Both the F-dolta-H a9d Fg(z) values proposed 1

are consistent with the values used for VANTAGE 5 fuel in I the Cycle 3 Licensing Submittal (Reference, ULNRC-1470 dated 3/31/87). The change in the Technical Specification i

! measured F-delta-H limit from 1.49 to 1.59 is supported by l l current DNB analyses., All DNBR margins and analyses arc as

described in the current callaway Plant Final Safety 1 Analysis Report (FSAR) and are unaffected by this change in F-dcltu-H.

! The chango in the Technical Specification F (z) limit from 2

2.32 to 2.50 has been evaluated to ensure t9at the cladding j l integrity and fuel molting at the "hot spot" are maintained ;

I within the applicabic safety analysis limits. All safety )

! analyses as described in the current callaway Plant FSAR are l unaffected by this change in F g(z). j

! The proposed changes to the Technical Specifications do not !

l affect the VANTAGE 5 design / safety bases used in the I VANTAGE 5 Reference Coro Report (WCAP-10444-P-A) or those l

contained in the Callaway reload safety evaluations.

i a) This change does not involve a significant increase in the probability or consequences of an accident previously evaluated. No changes are involved in accident initiators which would change the probability of an accident. The consequences of previously analyzed accidents remain unchanged sinco all DNBR margins are unaffected.

l 4 a

?

s' b) This change does not create the possibility of a now or different kind of-accident from any accident previously 7 ovaluated. This is besed on the fact that the method and manner ofcplant operation is unchanged.

, c) This chango does not involve a significant reduction in l a margin of safety. This is based on the fact that DNBR margins and safety analyses described in the Callaway Plant FSAR are unaffected,by the changes in F-delta-H and F9(z).

III. _ POSITIVE MODERA'!OR TEMPERATURE COEFFICIENT This evaluation has been performed to support the proposed

, Technical Specification change for the callaway Plant which l would allow a positive moderator temperature coefficient 1 (PMTC) of +5 pcm/'F to exist at power levels below 70%

nominal rated thermal power, with the maximum positive valuo

decreasing linearly to zero betwoon 70% nominal power and 100% nominal power. The use of PMTC supports reductions in fuel cycle costs by' reducing burnable absorber inventory, particularly for long fuel cycles which require a large number of burnable absorbers to control modcrator temperature coefficient at the beginning of a cycle. The Safety Evaluation performed in aupport of this amendment i request demonstrates that the sa.toty criteria are met for

the proposed PMTC limits.

a) This change does not involve a significant increase in

! the probability or consequences of an accident i previously evaluated. No changes arc involved in

! accident initiators which would change the probability i j of an accident. The consequences of previously

, analyzed accidents remain unchanged because the accidents and transients previously evaluated in the !

i FSAR which are sensitive to PMTC have been evaluated

. and/or reanalyzed and are not significantly affected by i the proposed change.

i b) This chango does not create the possibility of a new or different kind of accident from any accident previously 1

cvaluated. This is based on the fact that !

! implementation of PMTC does not involve any design i l changes to the fuel, reactor coolant system or i

engineered safety features which would create the possibility of a now or different accident than those l l

previously analyzed. i

! l c) This chango does not involve a significant reduction in i i a margin of safety. This is based on the fact that the l l affected safety analyscs have boon evaluated and it has i

! been determined that all applicable safety criteria are met with no significant adverse effects on analysis results.

2-l l

IV. RWST/ ACCUMULATOR BORON CONCENTRATION AND SAT SODIUM HYDROXIDE CONCENTRATION INCREASES The proposed changes would increase the RWST Technical Specification boron concentration range to 2350 ppm to 2500 ppm, and the accumulator Technical Specification boron concentration range will be increased to 2300 ppm to 2500 ppm. The purpose for the change is to allow for implementation of a positive MTC. In conjunction with these changes, the concentration of sodium hydroxide in the spray additive tank will he increased to a range of 31 to 34 weight percent.

The accident analyses reported in the FSAR were evaluated for the above increased concentrations. The results of the evaluation indicate that there is no adverse offcet on the FSAR results for any of the accident analyses as a result of increasing the RW6T and accumulator boron concentration operating bands nor as a result of increasing the spray additive tank sodium hydroxido concentration operating band.

a) This change does not involvo a significant increase in the probnbility or consequences of an accident previously evaluated. The proposed Technical Specification modifications involve changes in boron and sodium hydroxide concentrations in systems used to mitigate plant accidents and transients. The consequences of accidents and transients previously evaluated in the FSAR, including LOCA and non-LOCA events, have been ovaluated and/or reanalyzed and arc

not significantly affected by the proposed changes.

Iodine retention via the containment spray is assured, thuc offsito dose consequences are unchanged.

b) This change does not create the possibility of a new or

different kind of accident from any accident previously evaluated. The proposed Technical Specification changes would revise boron and sodium hydroxide concentrations in systems used to mitigate plant accidents and transients. No modifications to plant design or operations are associated with the changes with the exception of the operational requirement to switch to hot leg recirculation 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months following a LOCA, rather than 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . However, the 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months would allow sufficient time for the required operator actions and, therefore, the possibility of boron precipitation on fuel rods following a LOCA is not significantly affected, c) This change does not involve a significant reduction in l a margin of safety. The proposed revisions in boron and sodium hydroxide concentrations affect safety analysis inputs and assumptions. The affected safety analyses have been evaluated and it has been determined that all applicable safety criteria are met with no significant adverse effects on analyses results.

-3 -

SUMMARY

Dased on the above discussions, the amendment request does not involve a significant increase in the probability or consequences of an accident previously evaluated; does not create the possibility of a now or different kind of accident from any accident previously evaluated; and does not involve a reduction in the required margin of safety. Based on the foregoing, the requested amendment does not present a significant hazard.

1 l

l -<-

Attcchment 6 ULNRC- 1850 DRAFT FSAR CHANGES FOR CYCLE 4 section 15.4.6 pages 15.4-23, 24, 25, 26, 26a, 27, 28 Insert 1 i

l

CALLAWAY - SP

4. Overtemperature 4T turbine runback (at power)

5. Overtemperature AT reactor trip

6. Power range neutron flux - high, both high and low setpoint reactor trips.

This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1.

15.4.6.2 bnalysis of_ Effects and Consequqpcos To cover all phases of plant operation, boron dilution during Refueling, Cold Shutdown, Hot Shutdown, Hot Standby, Start-up, and Power modes of operation is considered in this analysis.

Conservative values for necessary parameters were used, i.e.,

high RCS critical boron concentrations, high boron worths, minimum shutdown margins, and lower than actual RCS volumes.

These assumptions result in conservative determinations of the time available for operator or system response after detection of a dilution transient in progress.

Dilution _During_ Refueling An uncontrolled boron dilution transient cannot occur during this mode of operation. Inadvertent dilution is prevented by administrative controls which isolate the RCS from the poten-tial source of unborated water. Valves BG-V-178 and BG-V-601 (or BG-V-602) in the CVCS will be locked closed during refuel-ing operations. These valves block all flow paths that could allow unborated makeup water to reach the RCS, Any makeup which is required during refueling will be borated water supplied from the RWST by the RHR pumps, pilution_ During, Cold _ Shutdown The following conditions are asaumed for inadvertent boron dilution while in this operating mode

a. Dilution flow is limited by a flow orifice in the RMWS to 150 gpm of unberated water,

b. An RCS water volume of 3400 ft' This is a cotiserva-tive estimate of the minimum active volume of the RCS and corresponds to the water level drained to mid-nozzle in the vessel while on one train of RHR.

c. All control rods (RCCAs) fully inserted, which is the normal condition in cold shutdown, and a critical boron concentration (Cn) o f 4* eta ppm . This is a con- l servative C with contl'ol rods i iserted and even allows for khe most reactive ro to be fully withdrawn. (gy Rev. OL-2 15.4-23 6/88

4 4

1 CALLAWAY - SP 3 '

4 centervativ: heren ucrtP ef I? pcm/pp.- l

p/,=*. The shutdown margin equal to 1 percent Ak/k, the minimum value required by Technical Specifications for i the cold shutdown mode. Combined with AssumptionX c 3 this gives a shutdown C B f Appm.

/fM i In the event of an inadvertent boron dilution transient while i in this mode of operation, the source range nuclear instrumen-tation will detect a doubling of the neutron flux by comparison 1

of the current source range flux to that of approximately

10 minutes earlier. Upon detection of the flux doubling, an c alarm is sounded for the operator, and valve movement to termi-nate the dilution and start boration is automatically initiated.

Under the conditions defined above, these actions will occur

! approximately 4 minutes after start of dilution. Valves l BN-LCV-1120 and E (isolation valves to the RWST) are opened to i supply 2000 ppo borated water to the suction of the charging

! pumps, and valves BG-LCV-112B and C (icolation valves in the i CVCS) are closed to terminate the dilution. These automatic actions are carried out to minimize the approach to criticality

)

and regain the lost shutdown margin. Action taken by the i operator is to terminate boration after regaining the required l shutdown margin and determine and correct the cause of the l

dilution transient.

3 Dilution Durina Hot _ Shutdown S

The following conditions are assumed for an inadvertent boron dilution while in this mode:

f' 1, a. The dilution flow rate is limited by piping system i

friction losses and the capacity of two makeup water

pumps to supply 260 gpm of unborated water.

C FO i' b. An RCS water volume of t,it? Vft'. This is a l

conservative estimste of the minimum active volume of j

the RCS, while on one train of RHR and with the RCS 3 . filled and vented.

]

c.

Allcontrolrodsfullyinsertedanda'-itica'[ boron concentration of 44eag ppm This is a conservative j l{ V*1** Of C a /100 When (buflown /Myin fr fok I ' '

_ence:v2tive beten :- r t5 :f -17 ;-- T;" l 5

p[, d- The shutdown margin equal to 1.3 percent ak/k, the minimum value required by Technical Specifications for j the hot ehutdown mode. Combined with AssumptionX c3 this g;ves a shutdown C 3 of + 9xppm. l j

a / 100 1

I Rev. OL-2 2

15.4-24 6/88

CALLAWAY - SP In the event of an inadvertent boron dilution transient while in this mode of operation, the source range nuclear instrumen-

-tation will detect a doubling of the neutron flux, automatically initiate valve movement to begin boration and terminate the dilution, and sound an alarm for the operator. Under the conditions defined above, these actions will occur approxi-mately minutes after start of dilution. ho operator action is requi ed to terminate this transient.

3'/a Dilution During Hot Standby The following conditions are assumed for an inadvertent boron dilution while in this mode

a. The diluton flow is limited to 260 gpm of unborated water (as in the previous case).

b. The RCS volume is 5800 ft'. This is a conservative estimate of the minimum active volume of the RCS with the RCS filled and vented and one RCP running. / s> -

p)h en shuNewn /M In

c. A a m N oron concentration of pm This is l Af b a conservative C assuming all control rods are fully

L t '& = ; inseked,

- d. A 0;r.;;rvr.tiv: bcrc: ;lOrtP Of -12.0 ?CE/ppr 0/,4% The shutdown margin equal to 1.3 percent ak/k, the minimum value required by Technical Specifications for the hot standby mode. Combined with AssumptionX cp**+

war this gives a shutdown C B f* APP"' l

/f@

In the event of an inadvertent boron dilution transient while in this mode of operation, the source range nuclear instrumen-tation will detect a doubling of the neutron flux, automatically I

initiate valve movement to begin boration and terminate the dilution, and sound an alarm for the operator. Under the con-ditions defined above, these actions will occur approximately l 27-4 minutes after start of dilution. No operator action is l required to terminate this transient.

DJ_1u t i_oILDu ting _S t a r t-up l

In this mode, the plant is being taken from one long-term mode of operation, Hot Standby, to another, Power. The plant is maintained in the Start-up mode only for the purpose of start-up testing at the beginning of each cycle. During this mode of operation, rod conttol is in manual. All normal actions required to change power level, either up or down, require operator initiation, The Technical Specifications require an Rev. OL-2 15.4-25 6/88 L

CALLAWAY - SP available trip reactivity of 1.3 percent AK/K and four reactor coolant pumps operating. Other conditionn assumed are:

a. Dilution flow is the maximum capacity of two centri-fugal charging pumps with the RCS at 2250 psia:

approximately 245 gpm.

b. A minimum RCS water volume of 9700 ft'. This is a very conservative estimate of the active RCS volume, '

minus the pressurizur volume.

c. 3CNJct1~ /

--c Ecr the-purpecc cf--thic nclycin, the criticci C c".ed te be 2100-pr fte: seccter trip. Thic 0:ic 1ue ir conccrvativc cad cnvcbepc th: lcret eccc, cetuci critical bcron concentretica cc:recpetwling te hot rcrc powcr, ;11 redo cut, nd ac c cc ' loc

. cccred ir c conee:vetive, cenetent bere: >crth cf -- '

12." pcm/rpn

=-! The nec sc ry chutdown margin, ac scquircd by the Technical Epccificatienc, ic previd:d-by the centrci

-- redc after rcccter trip Mc credit is t; hen-for thc scerc aegative-reactivity inscrted-by thc control rode which ic abcvc the chutdcu: : rgi:-

This mode of operation is a transitory operational mode in which the operator intentionally dilutes and withdraws control rods to take the plant critical. During this mode, the plant is in manual control with the operator required to maintain a very high awareness of the plant status. For a normal approach to criticality, the operator must manually initiate a limited dilution and subsequently manually withdraw the control rods, a process that takes several hours. The Technical specifications require that the operator determine the estimated critical position of the cor, trol rods prior to approaching criticality, thus assuring that the reactor does not go critical with the control rods below the insertion limits. Once critical, the power escalation must be sufficiently slow to allow the operator to manually block the source range reactor trip after receiving P-6 from the intermediate range (nominally at 10' cps). Too fast a power escalation (due to an unknown dilution) would result in reaching P-6 unexpectedly, leaving insufficient time to manually block the source range reactor trip. Fa' lure to perform this manual action results in a reactor t.ip and immediate shutdown of the reactor.

Rev. OL-2 15.4-26 6/88

CALLAWAY - SP

u;v: :, in th: :v:nt :f : a.;1 r. d ;;r:::h t: ;ritice,11ty or diluti:n during p:Ucr ::: 1 tion chil: ir th: St::t up

de, the pl nt et:tu: i: uch th:t minia:1 impect wirl r : ult. The pl:nt uill :1:uly ::::let: in p:u: t: : :::t:r trip :n th; ;;u:r ::ng; high n: ts:n flun leu ::tp int

'n:ninelly 25 per::nt RT"} After reactor trip, there is approximately 4H> minutes for operator action prior to return to criticality. The required operator action is the opening l of valves BN-LC -112D and E to initiate boration and the closing of val s BG-LCV-112B and C to terminate dilution.

40 Dilution Durino Full Power Operation The plant may be operated at power two ways: automatic T,yg/ rod control and under operator control. The Technical Specifications c

P i

5 f

P l

I r

Rev. OL-2 15.4-26a 6/88

CALLAWAY - SP require an available trip reactivity of 1.3 percent AX/X and four reactor coolant pumps operating. With the plant at power and the RCS at pressure, the dilution' rate is limited by the capacity of the centrifugal charging pumps. The analysis is pergformed assuming two charging pumps are in operation even thoughrnormal operation is with one pump. Conditions assumed for this mode are:

a. Dilution flow from two charging pumps is at the mani-mum at an RCS pressure of 2250 paia (approximately 245 ypm) when the reactor is in manual control. When in automatic control, the dilution flow is the maximum letdown flow (approximately 120 gpm).

b. A minimum RCS water volume of 9700 ft'. This is a very conservative estimate of the active RCS volume, minus the pressurizer volume,

c. .~fN.fE/7~ /

c. rer the purpoas cf-this analysia, the transient initial heren ::ncentration i :::u.;d to be 1000 ppe Thi: value i: cons;rvativ: and :nv;1ep; th; U: ret-:::c, ::tu:1--:riti 1 beren concentration cerra:pending tc-het full pcuer, all-red: Out, and n X:non ?.1:: :: rmed i con ervative, 0: net:nt beren worth of 13.0 p =/p7r

d. Th n :::::ry chutdeun rargin, 20 requir:d by the T= Ente =1 sp::iricatione, 10 provided by the centrol red: :fter-re::ter-trip, N: :::dit in t:9:n for th ex:::: neg:tive reactivity in:crted by th: centrol r:d: uh:ch 1: & cvc th: hutd:= :rgin f of learf 34 With the reactor in manual cor trol and no operatcr action taken to terminate the transie nt, the power and temperature rise will cause the reactor tc ret.ch the Overtemperature AT trip setpoint resulting in a 1 eactor trip. After reactor trip there is sper+*4-metely 15.75 Uminutes for operator action l prior to return to criti-cality. The required operator action in the opening of valver. BN-LCV-ll2D and E and the closing of valver BG-LCV-ll2B and C. The boron dilution transient in this case is essentially the equivalent to an uncontrolled rod withdrawal at power. The maximum reactivity ineartion rate e

gg for a_ boron dilution transient is conservatively estimated to i be evilpem/sec and is within the range of insertion rates l i analyzed for unco @ rolled tod withorawal at power. It should

] be noted that prio* to reaching the Overtemperature AT reactor i trip, the operat.. 'ill have received an alarm on overtemperature tT ianOvergtemperatureATturbinerunback.

With the reactor ...tomatic red control, the pressurizer 1 level controller wiAl limit the dilution flow rate to the i maximum letdUown rate, approximately 120 gpm. If a dilution

! rate in excess of the letdown rate is prosent. the pressuriser l

1evel controller will throttle charging flow down to match the letdown rate.

4 1

! Rev. OL-2

[ 15.4-27 6/88

c i

i CALLAWAY - SP l

b. Red insertion limit - low-low level alarm if i insertion continued after item a l

. Axial flux difference alarm (A! outside of the tr.vget l band). :

Given the many alarms, indications, and the inherent slow

(" 'll i process of dilution at power, the operator has sufficiont time :

for action. For example, the operator has at least +0"minutes i from the rod insertion limit low-low alarm until 1.3 p::::nt l 1",~ i; in:::t:d at beginning-of-life. The time would be significantly lo ger at end-of-ll'le, due to the low initial i boron concentra on. - . . . -.._----.. _ ,... -- - - - . . . .

dufalen Nrjin it lor+

The above resulte demonstrate that in all modes of operation, i

an inadvertent boron dilution is precluded or responded to by !

automatic functions, or sufficient time is available for oper- I ator action to terminate the transient. Fol") wing terminati,n ;

of the dilution flow and initiation of boration, the reactor is I in a stable condition with the operator regaining the required i shutdown margin.

i 15.4.6.3 Conclusions f The ::: ult: pre::nt:d pr:tice:1y ch: 1 th:t 3hadvertent boron !

dilution events are precluded during refueling and automatically terminated during cold shutdown, hot shutdown, and hot standby '

modes. Inadvertent boron dilution events during start-up er .

power operation, if not detected and terminated by the i operators, will result in reactor trip. Following reactor trip, there is ample time available for the operators to i terminate the dilution prior to a return to criticality, t r

15.4.7 INADVERTENT LOADING AND OPERATION OF A TUEL ASSEMBLY IN IMPROPER POSITION !

I 15.4.7.1 IdentiLigation of causes and_ Accident Desiription

l Fuel and core loading errors that can arise fro.n the inad-l vertent loading of one or more fuel assemblies into improper p

positions, loading a fuel rod during manufacture with one o- >

I more pelleta of the wrong enrichment, or the loading of a full l j

! fuel assembly during manufacture with pellets of the wrong

! ent'.chment will lead to increased heat fluxes if the error .

l results in placing fuel in core position. calling for fuul of I lesser enrichment. Also included among possible core loading l errors is the inadvert6nt loading of one or more fuel assem- i blies requiring burnable poison rods into a new core without f burnable poison rode.

Any error in anrichment, beyond the normal manufacturing i tolerances, can cauce power shapes which are more peaked than j those calculated with the correct enrichments. There is a ;

i l i l w . on-1 l

[ 15.4-28 6/87 !

t __

p, ,- . . -

, e ,

o..

c

)- p e i. ,

l-~ ,

., \ ~

INSERT-1

). -

4 An initial maximum critical boron concentration, .

corresponding to the rods inserted to the insertion limit,,,

is assumed to be~1800 ppm. The minimum change'in boron l concentration frsm.thic initial condition to a hot zero power critical 7'eondition with .all rods inserted is ' assurwd to be 300 ppm.-- Full rod insertion, minus the'most reactive, stuck rod, is assumed to occur due to reactor trip.

,o i

t

,

ML20205N395

Text

3.2.3 Fhshallbelimitedbythefollowingrelationship

SUMMARY

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