Text

Rev 0 to Catawba Unit 2,Cycle 6 Colr
	ML20126F026
Person / Time
Site:	Catawba
Issue date:	11/22/1991
From:	Clark R, St Clair R; DUKE POWER CO.
To:	;
Shared Package
ML20126E996	List: ML20126E995; ML20126F014; ML20126F026;
References
	CNEI-0400-14-01, CNEI-0400-14-R00, CNEI-400-14-1, CNEI-400-14-R, NUDOCS 9212300086
	Download: ML20126F026 (39)
	v • d • e

CNEl Nm.14 Page i of 15 Rev um Catawba Nuclear Station COLR b

Catawba Unit 2 Cycle /

i Core Operating Limits Report November 22,1991-Duke Power Company Piepared by: TMP l dL4 Checked by: %< h 'UM4 C v

Approved by:

A MC 5-L 9212300086 921215 PDR.ADOCK 05000414 8 '.31 p

PDR

CNEl NIH 14 Page 2 of 15

, (>

Rcv. ot XI Catawlia 2 C,sclef Cure Operating Liinits 1(eport itEVISION LOG Revtsion Effective Date Effective Paces Original issue 22 November 1991 Pages1 15 Y

i i

h 8 132

.. _..~

i I

CNiii.(uno.la Page)ofI?

b Rev.oto Cataw ha 2 C, tele / Core Operating Lirnits Heport i

i l

l 1.0 Core Ooerniine 1 imits Itcoort 6

)

His Core Operatine Limits Report. (COLR). for Catawba Unit 2. Cycle / has been prepared in acco* tance with the requirements of Technical Specification 6.9.1.9.

I The Tetnnical Specifications aticcted by this report are listed below:

3/4.1.1.3 Moderator Temperature Coeffielent i

3/4.1.3.3 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod insertion Limit 3/4.2.1 Axial Flux Difference Ji4.2.2 Heat Flus Hot Channel Factor j

3/4.2.3 Reactor Coolant Spiem Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor l

v 8-133

1 1

t CNt!! tuno.14 Page 4 of 15 h

Rev, m l i

Catawha 2 C,scle/ Cure Operating 1.imits Heport I

'0 Ooeratine 1.linitt

{

The cycle specific parameter limits for the specifications listed in section 1.0 are presented in the following subsections. These limits have been developed using NRC approved methodologies specified in Technical Specineatior 6.9.1.9.

i 2.1 Moderntor Temnerature ('oef ficient iSnecillention.14.1.1.3) l 2.1.1 The Moderator Temperature Coefficient (MTCi Limits are:

l The MTC shall be lew positive than the limits nown in Figure 1. The BOC/ARO/liZP MTC shall be less posidve that u.7

10 4 AK/KrF.

The EOC/ARO/RTP MTC shau be less negatise that.4.1

10 4

.iK/K/ F.

3 2.1.2 The MTC Surveillance Limit is:

The 300 PPM /ARO/RTP MTC should be less negative than or equal to 3.2

10 4 AK/K/ F.

'Vhere:

DOC stands for Beginning of Cycle ARO stands for All Rods Out liZP stands for liot Zero (nermall Power i

EOC stands for End of Cycle RTP stands for Ruted Thermal Power 1

8 134

CNEl 0401bi4 Page5ni15 Res, o181 6

Catawha 2 C3cle / Cure Operating Limits Report I

10 0.9 --

4 3

.M -

Unacceptable Operation 0

o g 0.7

=.

k 0.6 --

'O

]0.5--

Acceptable Operation E

3 0.4 --

03 -

e 0.2 -

t3

= 0.1 -

0 0

10 20 30 40 50 60 70 80 90 100 Percent of 11ated Thermal Power Figure I

\\1oderator Temperature Coefficient Versus Power Level 8-135

2

.i i

CNEl-@ W1 la i

Pagenosl$

h Rev.one j

Catawba 2 Cycle / Cure Operating Limits Report j

i I

l 1

l 2.2 Chutdow n Rod Insertion 1.ltnit (Enecillention 314.1.3.8) 4 2%>

2.2.1 The shutdown rods shall be withdrawn to at least.22(steps.

2.3 Control Rod insertion 1.imits 'Enecillention 3/4.1.3.61 1

1 i

l_

2.3.1 ne contml rod banks shall be limited in physical insertion as shown in 1-Figure 2.

1 i

}

2.4 uini illus liifference iKnecillention.t!4.2.11 t

i-

.i 4

De AXlAL FLUX DIFFERENCE ( AFD) Limits are provided in Fi '

1 l

2.4.2 The target band dui..

load oper "

upplicable for l

Catawba 2 Cye:e 5.

l

.2Ar3 1e minimum allowable power level for Base Load Operation

FL101, l

is not applicuole for Catawba 2 Cycle 5.

l i

i i

f 2.4.1 The Axial Flux Difference (AFD) Limits are provided in Figure 3.

(AFD Limit)COLR is the negative AFD limit from Figure 3, negadve I-

{:

(AFD Limit)COLR is the positive AFD limit from Figure 3.

pigy, 4

?

4..

y 4.

h

CNEl-(Wrl 14 Page 7 of 15 Rev ofn Calauha 2 C,5cle 5 Core Operatitig 1.imits Report Fully Withcrawn \\

(27,4 % 225)

(77.8 % 226) y 220 ~\\\\\\

~

200 CANKB

~

I80 (01163)

(100 % 161)

60 g,p fy, h

\\

ta d4%

~

40 BANK C Ju s cef

3 4

=

3

20

~

2 ET b

100

~

2

?

30 x

I B At K O 6

~

9 50

=

3 (0%47)

~

40

~

20 (30%0)

Fully Insertea \\

l I

I I

l l

0 20 40 60 80 10 Relative Power (Percent)

Figure 2 Control Rod Bank Insertion Limits Versus Percent Rated Thermal Power t

8 137

Insert 4 Cataulia 2 Cycle 6 Core Operating 1.imits Report l

Fully Whhdrawn (29'%,. 230) (Matimutn=230)h (80'3,230) 230

* ~ ~ * * * * ' " " * * *

~~{~~*"***"~~~~***"*"***"

2'O -#~*"~~~"""~

' Fully Withdrawn (Minimum =222) 2(X) -

i 7 ggo __

II ANK 11 W

R

( 100'7c,161) j INI ~~ (O'4, I M )

G

$ 140 ?

6 II ANK C g 120 -

i a

[100-8 l

9 80 -

llANK D jj 60 -

an.. (O'X, 4 H i 20 --

(39,g, o, Fully inseneds 0

l

A

i 0

10 20 30 40 50 N) 70 80 90 100 l

l'ercent of Rated Thermall'ower l

Figure 2 Control Rod Bank Insertion Limits Versus Percent of Rated Thermal Power 8-138

CNE14Wu 14 Pageaol!!

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Catawba 2 Cycle) Core Operating Limits Heport IN i

t a

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10 20 30 40 A taal Flut Differevue t % Dette 16 Figure 3 -

Percem of Rated Dermal Power Versus Axial Flux Difference Limits 4

8139

1 i

CNEl Mn11 1

Parcyof 14 Rey, im l

Catawha 2 C3cle 5 Cure Operating Lirnits l{cport J

j 25 llent Flut flot Channel Factor Fq(Z)(Specification N4.2.2) i, i

F(

s. F""

Kf Z) for P > 0.5 n

1

()

l P

F (Z) s F" '

KfZ) for P s 0.5 n

4 o

U.5 Thermh(Power ycicewsh where: P=

Rated Therm l Po It ferf 5 i

l 2.5.1 F"" = 2.32 o

i 2.5.2 K(Z)' provided in Figure 4.

j l

2.5.3 % '(Z) values are provided in Figures hrough 7.

f 2.4 Base load W(Z)'s are not applict.ble for Ca wba 2 Cycle 5.

i

.{.

?

.,L 8 140

t

=.:.

Insert 5 -

Catawba 2 Cycle 6 Core Operating Limits Report 2.5 llent Flut flot Channel Factor.FQfX.Y.2) fSnecinention 3/4.2.2)

RTP 2.5.1 F

= 2.32 g

2.5.2 K(Z) is provided in Figure 4 for Mark-BW fuel.

2.5.3 K(Z) is provided in-Figure 5 for OFA fuel.

ne following parameters-are required for core monitoring per the' Surveillance Requirements of Specification 3/4.2.2:

D

[F (X,Y,Z)]OP = O(X,Y,Z)

M (X,Y,Z)

F g

. Q 2.5.4 9

where IF (X,Y,Z))OP =

' cycle dependent maximum allowable design peaking factor which ensures that the F (X,Y,Z) limit will be-Q pr(served for operation within the LCO limits.

[F (X,Y,Z)]OP ncludes allowances for calculational and i

9 measurement uncertainties.

F (X,Y,Z) = the design power distribution for F. F (X,Y,Z)is provided Q

in Table 2 for normal operation and table 2A for power escalation testing during initial startup.-

M (X,Y,Z) = the margin remaining in core location X,Y,Z to the LOCA Q

limit in the transient power distribution. M (X,Y,Z) is Q

provided in Table 3 for nomial operation and table 3A for power escalation testing during initial startup.

UMT = Measurement Uncertainty (UMT = 1.05).

MT = Engineering Hot Channel Factor (MT = 1.03).

TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio.

of 1.02.-

g,Y,Z)]OP s'the parameter identified as F (X,Y,Z)in DPC-NE '

NOTE: [F X

i g

2011PA.

8 141

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

)

Insert 5 - continued Catawha 2 Cycle 6 Core Operating Limits Report F (X,Y,Z)

M (X,Y,Z)

C 2.5.5 lF (X,Y,Z)lRPS.

where [F (X,Y,Z))RPS =

cycle dependent maximum allowable design peaking factor which ensures that the centerlig(fuel melt lim will be preserved for all operation. [F X,Y,Z)]RPS includes allowances for calculational nd measuremer.c uncertainties.

F (X,Y,Z) = the design power distributions for F. F (X,Y,Z)is provided Qq in Table 2 for normal operation and table 2A for power escalation testing during initial startup.

M (X,Y,Z) = the margin remaining to the CFM limit in core location X,Y.Z C

from the transient power distribution. M (X,Y,Z)

C calculations parallel the M (X,Y,2) calculations described in Q

DPC-NE-20llPA, except that the LOCA limit is replaced with the CFM limit. M (X,Y,Z) is provided in Table 4 for C

normal operation and table 4A for power escalation testing during initial startup.

~

UMT = Measurement Uncertainty (UMT = 1.05).

MT = Engineering Hot Channel Factor (MT = 1.03).

TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio j

of1.02.

NOTE: [F (X,Y,Z)]RPS is similar to the parameter identified as (X,Y,Z) in DPC-NE-2 11PA except that M (X,Y,Z) replaces M (X,Y,Z).

C Q

2.5.6 KSLOPE = 0.078 where KSLOPE = Adjustment to the K valu{from OTAT required to compensate 1

for each Ir/c that [F (X,Y,Z)]RPS exceeds it limit 9

8-142

.... -... - ~ -

i CNEl 04(n-14 Page 10 of 15 l

Rev.i m 3

i e

4 I-a l

?

j

- Cataw ba 2 Cycle 5 Core Operating Limits Report e

1.2 i

ke lace.

1 (0.0,1.00) l l(6.0.1.00)l 1-(10.8,0.94) yj&

In 5er$~

l 0.8 (o

o

@ 0.6 (12.0.0.647) 4

[

0.4 i

i a

l' O2 b

l-0 1

-2 3

4 5

-6 7

8 9

10-1 12 1

Core Height (ft)

. ~.

Figure 4-K(Z). Normalized F tZ) as a Function of Core Height.

Q I

8-143 4

, ~...

r

,,~,,evr~7

-e..

+

Insert 6 Catawba 2 C3 cle 6 Core Operating Limits 1(eport n

1.2 l

l(0.0,1.00)

(8.0.1.00) 1 (10.8,0.94) 0.8 4

(12.0.0 M7)l h 0.6 i

l 0.4 4

)

~

O.2 0

i 0

1 2

3 4

5 6

7 8

9 10 11 12 Core lleight (ft)

Figure 4 K(Z), Normalized F (X,Y,Z) as a Function of Core Height for Mk-BW Fuel Q

8-144

y 1

3 i

J

+

3 I

Insert 6 - continued Catawba 2 Cycle 6 C6re Operating Limits Report i,

f i

i 2

1.2 4

4..

I

)(0.0.1.00)

(6.0.1.00)

(10.8.0.941 j

g -.

1p-1 j--

0.8 i

k E

i' E o.6 (12.0.0.647 d

.t 4

5 I

n.s 4

i i

l 0.2 i.

4 i~

j g

0

-1 2

3 4

5-6 7'

-8 9

10' 11 --

12 Core Height (ft) s i

a c

-1.

Figure 5 i

i K(Z). Normalized F (X,Y,Z) as a Function of Core Height for OFA Fuel i

Q a

t' 1i i

'8-145' 1I-

1 i

i i

)

CNEl G10014 1

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Rev.000 j

Catawba 2 Cycle 5 Core Operating Limits Report 4

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CNEl-04t o-14 Pape 12 of 15 -

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Catawha 2 Cycle 5 Cure Operating Limits Report

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i

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8 10 12 ii; bestem CCRE HEIGHT (TEET) top 4

1 l

Figure 6 1

Catawba Unit 2 Cycle 5 RAOC WlZ at ag]g.) LWD /MTU

~

Top and Buttom (M eniuded as per Tech. Spee.4.2.2.2.G '

4

- - +

..m.,T m-.-.-.,

., ~ ~,.

,,w,

,,.m,,m3

1-d 4

i CNEbNm.14 s

1

- Page 1,1 on 15 '

Rev. iHN)

I Catawba 2 C.scle 5 Core Operating Limits Report i

XII T EOL f

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t il

'18(El l

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4 6

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12 dest==.

c::RE mcIcar tru:n sg i

Figure 7 Catawba Unit 2 Cycle 5 RAOC W(Zi at 13000 MWD /MTU l

8-148 Top and Bottom 15?c excluded as per Tech. Spec. 4.2.2.2.Gl

.r 4

r-r ww+

,--v--,-a m

o,..m-+.er 3-w w ~-e v vvw i-t rr-+ m

-+.*--s w r,riv

=

=?.+ne

i 4

i CNEl 0400-14 Page 14 of l$

1

-Rev.In1 Catawba 2 Cycle 5 Cure Operating Limits Report l

i i

{.

- 2.6 R '9 Flow Rnte and Nuciene Fnthnlov Rise Hot Channel Fnctor. -

AH (Spec cation N4.2.3) 2 1

j FN 3g R=

i FRTPAH *

+MFaa (1 P))

3 egl ace weg i

Thermal Pow A re,-f 7 1

- where: P=

i Rated Thermal Po er i

2.6.1 FRTP g = l,49 -

2.6.2 MP a = 0.3 3

k 2.6.3 The Ace table Operation Regio.n from the mbination of Reactor j

Coola System total flow and R is provided in igure 8.

T.

i.-

6 4

i f.

i I

a 4

s 8-149-

d i

4, l'

Insert 7 Catawba 2 Cycle 6 Core Operating Limits Report l

2.6 Nuclear Enthalov Hise 1101 Channel Factor. FaygfX.Y.2) (hcification 3/413)

The following parameters are required for core monitoring per the LCO Requirements of Specification 3/4.2.3:

~

1 2.6.1

[ Fall (X,Y)]LCO = MARP (X,Y)

1.0 + RRii (1,0 - P)

{

where (MARP(X,Y)) = Catawba 2 Cycle 6 Operating Limit Maximum Allowable-l Radial Peaks. (M ARP(X,Y)) is provided in Table 1.

{-

Thennal Power E " Rated Thermal Power i

The following parameters are required for core monitoring per ' the Surveillance l

Requirements of Specification 3/4.2.3:

D 3g(X,Y)]SURV, 3g(X,Y)

Magj(X,Y)'

F

.g 2.6.2 [F 5

L SURV =

cycle dependent maximum allowable design peaking i

where [F3g(X,Y)l factor which ensures that the F g(X,Y) limit will be A

l pr(served for operation within the LCO limits.

l

-[FAH(X,Y)]SURV includes allowances for calculational i-and measurement uncertainties.

D D

FAH(X,Y) = the design power distribution for FAH FAH(X,Y)is provided 4

in Table 5 for normal operation and table 5 A for power j

escalation testing during initial startup.

MAH(X,Y) = the margin remaining in core location X,Y to the Operational

[

DNB limit in the transient power distribution.- MAH(X,Y) is.

provided in Table 6 for normal operation and table 6A for power escalation testing during initial startup.

UMR = Uncertainty value for measured radial peaks,(UMR = 1.04).

4 -[

TILT = Peaking penalty that accounts for allowable qu'adrant power tilt ratio i

of1.02.

8 150

' vv i

y

,em-..a yU, p

-,y

,,ym y, e-g sw g

_-.ny-awe gy,

.c es,,--,y fy

,~

p m v

q.,

,y.,-

ysw,

4 l

/

4 Insert 7 - continued Catawba 2 Cycle 6 Core Operating Limits Report NOTE: [F lX'y))SUR\\r s the parameter identified as F ;AX L

M (X,Y)in DPC-NE-All 3j 201 IPA.

2.6.3 RRii = 3.34 where RRif = Thermal Power reduction required to compensate for each 1% that l

FAli(X,Y) exceeds its limit.

4 2.6.4 TRil = 0.04 where TRif = Reduction in OTAT K setpoint required to compensate for each 19 1

that Fall (X,Y) exceeds its limit.

8-151

o 4

Insert 7 - continued Cataulm 2 Cycle 6 Core Operating I.imits Report i

l Table 1. Maximum Allowable Radial Peaks (M ARPs)

Core lleicht 1.1 Axial Peak 1.2 Axial Peak 1.3_ Axial Peak 1.4 Axial Peak W

MARP MARP MARP MARP 0.12 1.5809 1.6266 1.6722 1.7113 1.2 1.5806 1.6259 1.6677 1.7085 2.4 1.5836 1.6265 1.6663 1.7025 3.6 1.5859 1.6263 1.6635 1.6960 4.8 1.5871 1.6240 1.6571 1.6751 6.0 1.5878 1.6196 1.6470 1.6303 j

7.2 1.5864 1.6130 1.6265 1.5848 2

8.4 1.5781 1.5956 1.5773 1.5327 9.6 1.5655 1.5612 1.5208 1.4815

)

10.8 1.5459 1.5152 1.4717 1.4292 12.0 1.5133 1.4693 1.4274 1.3878 l

Core Heicht 1.5 Axial F.rak 1.6 Axial Peak 1.7 Axial Peak 1.8 Axial Peak 4

W MARB MARP MARP MARP l

0.12 1.747~

1.7331 1.7054 1.6438 j

1.2 1.7433 1.7029 1.6789 1.6193 l

2.4 1.7126 1.6616 1.6433 1.5869 3.6 1.6735 1.6211.

1.6011 1.55N 4.8

!.6313 1.5811 1.5622 1.5121 i

6.0 1.5868 1.5415 1.5238 1.4763 l

7.2 1.5378 1.4913 1.4766 1.4344 8.4 1.4886 14450 1.4296 1.3880 9.6 1.4399 1.4013 1.3882 1.3490 10.8 1.3883 1.3526 1.3433 1.3081 12.0 1.350(

l.3140 1.3078-1.2749 4-I Core Heicht 1.9 Axial Peak 2.1 Axial Peak j

MARP MARP 0.12 1.5839 1.5401 1.2 -

1.5624 1.5154 2.4 1.5328 1.4801 3.6 1.5013 1.4395 4.8 1,4626 1.4030 6.0 1.4291 1.3619 7.2 1.3920 1.3271 8.4 1.3485' l.2824 9.(

l.3126 1.2501 10.8 1.2726 1.2091 12.0 1.2443 1.I890 8-152 4

d t

CNEl St(X)-14 y

Page 15 of 15 s

Rev.000

N Catawba 2 Cycle 5 Core Operating Limits Report i

'\\

N Penaities 10.1% for uncatected feeowater ventun fouling and measurement uncertainties of 2. % for flow incere measurement of F' hare inctuded in this figure.

ano 4.0% 1.

1 1

40.0 1

l 39.5 Permissible p,,g gy gg,g O eration P

op,,,,,,,

9 l* "

Region I

29.0 -

g g

(1,000, 38.760) i

$. 38.5 - Restncted Operati n Region (Power 198Y TP) 3,,,

(0.994 38.372)

Restncteo Operation Region fPower 196% RTP tu 38.0 (0.988, 37.985)

o Restncteo Operation Region (Power 19 RTP) d (0.982. 37.597) g 37.5 - Restncted Operation Region (Powe/

r 192% RTP)

(0.977. 37. 10) 37,o-Restncted Ocoration Regio (Power 190% RTP) o (0.971. 36.822) 36.5 -

i 26.0

[

0.93 0.94 0.95 0.96 0.97 0.98 0.99 1.00 1.

1 1.02

/

R. FOUR LOOPS IN OPERATION Figure 8 RCS Flow vs. R Four I. oops in Operation I

8-153

. - - - - - - _ _ =

4 8.3' Changes to the Catawba FSAR.

i n

8-154

-. ~.

_ _; -_.. u _;_._ _.

i k

2 4

1 li Catawba Nuclear ' Station -

- Appeaux 6.; Chapter 6 Table ana figures l

~

Notes to Table 6 77 5-Con'=inment isolation Valve and Aetn= tion Data -

i i.

Notes:

3

1. Valve arrangements are shown in Figure 6112.

t

2. Dennition of Actuation Signals l

S Safety injection Signal (T signal also activated by S signal)--

i 1

T Con'=== at isolation Signal (Phase A conta==ent isolation) t j

P Cont===ent High High Pressure Signal Phase B containment. isolation) f 0

3. Valve Type Abb... dcas j

0 GL Globe-

.CRR5e5 rndn 'sfeetm //4c:. - /so/4M/on %[

s

-0 SW Swing Check 1

j --

0

. GT Gate 1

l 0

CK Check i-0 RV Relief l

0

-DS Double Seal -

(-

0 FG Flange i

0 PG Plug 1

O BF Butterfly l

]

O DP_ Diaphragm i

0 SV Safety j

L i

0 SC Stop Check

4. Symbols:

}

Valve Position Abbreviations 1-

]-

O - Open

-C' Closed j

A-Automatic -

n R

Remote Operauon i

M Manual Imal Operation 4

LC - 1.ocked Closed p

ClO Closed prior to S' ump or Hot leg Rectreulation: Open after Sump or Hot leg Recirculation LO.. Locked Open

'Al Fails As is 4

j.:

Actuator Tme E'

. Motor i Power Source. Electncity).

i.

I 8-155 (01 OCT 1991):

+

. - ~ -

,a-.,,.

..._,_...,-.a,.

I, ISA Heretivity end Power Distribution Anomdies Catawba Nuclear Station.

1 Dilution Flow Rats thuimum I

in the absence of Do rate restrictions, the dilution flow rate assumed to enter the RCS is greater than or i

equal to the olumetric flow rate of both reactor makeup water pumps. In a dilution event, these I

pumps are assumed to deliver unborated water to the suction of the centrifugal charging pumps. Since I

the water delivered by these pumps is typically colder than the RCS inventory, the unborated water I

expands within the RCS, causing a given volumetric flow rate measured at the colder temperature to 1

correspond to a larger volumetric dilution Dow rate within the RCS. This density difference in the 4

i I

dilution flow rate is accounted for in the analysis.

I Results 1

1 The calculated sequence of events is shown in Table 15 23.

1 Dilution Durinc Modes in which the BDMS is Reauired (Modes 3 6) 1 During Mode 6 an inadvertent dilution from the Reactor Makeup Water System is prevented by I

cdministrative controls which isolate the RCS from potential sources of unborated makeup water. The I

results presented in Table 15 23 for this mode are for an assumed dilution event, for which no mechanism i

1 or flow path has been identified. For Modes 3 6 with the BDMS operable, the results presented in i

1 Tcble 15 23 show that there is adequate time to reach the BDMS alarm setpcint, stroke closed the valves I

to isolate the source of unborated water, and purge the unborated water already in the CVCS piping, I

before the shutdown margin is exhausted. For Modes 3 6 with the BDMS inoperable, the results I

presented in Table 15 23 show that, with limitations on flow rates from potential sources of unborated 1

wcter, there is adequate time for the operator to determine the cause of the dilution, isolate the source of l

i i

unborated water, and initiate reboration before the shutdown margin is exhausied. In accordance with 1

Reference 11, adequate time is judged to be at least 15 minutes for Modes 3 5 and at least 30 minutes for 1

Mode 6. The results presented in Table 15-23 are for the dilution flow rates which, assuming the boron I

concentration ratios are at the reload safety analysis limits, give exactly these operator response times.

Flow rates are restricted, through Technical Specifications and administrative controls, to values which are 1

I less than these analyzed flow rates, thus in practice giving even longer operator response times. Additional 1

margin is provided by the fact there is typically margin between the assumed boron concentration ratio for 1

a given mode and the actual corresponding concentration ratio for the reload core.

Dilution Durinc Startun (Mode 2)

This mode of operation is a transitory mode to go to power and is the 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 For a normal approach to criticality the operator must manually initiate a limited dilution and stztus.

subsequently manually withdraw the control rods, a process that takes several hours. The plant Technical Specifications require that the operator determine the estimated critical position of the control rods prior to approaching criticality thus assuring that the reactor does not go critical with 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 105 cps). To 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. Failure to perform this manual action results in a reactor trip and immediate shutdown of the reactor, allowing suflicient time prior to a loss of shutdown margin for the operator to termmate the dilution event.

However, in the event of an unplanned approach or dilution during power escalation while in the startup mode, the plant status is such that muumal impact will result. The plant will slowly escalate in power to 8-156 15-78 (01 OCT 1991)

~

i 4

l i

Catawba Nuclear Station Appendix 15. Chapter 15 Tables and Figures i

i Table 15-23 (Page 3 of 4). Time Sequence of Events for incidents Which Cause Reactivity and Power Distribution Anomalies i

Time -

Accident Event (sec.)

i 1

4b, Dilution during hot Dilution begins 0

]

I shutdown (BDMS -

I inoperable) 1 Iligh Oux at shutdown alarm setpoint reached 1816 l

1 Operator terminates dilution

< 2716 l

1 Sa. Dilution during cold

' Dilution begins _

0 i

i shutdown (BDMS 3

1-operable)

To6 i

1 BDMS setpoint reached M9' I

Dilution source isolated Br F3I a

1 Borated water reaches core

,s-885 4 17/

j i

. Sb. Dilution during cold Dilution begins 0

I shutdown (BDMS 1

inoperable).

152p 18t6 i

1 High flux at shutdown alarm setpoint reached i

1 Operator terminates dilution

.s-2?ig 4272o l

6a D:1ution during refueling Dilution begins 0

I (BDMS operabic) l 1

BDMS setpoint reached 1024 1

l 1

Dilution source isolated 1049 l

1 Borated water reaches core

< 1267 I

6b. Dilution dudng refueling Dilution begins -

0

-1 (BDMS inoperable) 1 High Dux at shutdown alarm setpoint reached 3441 1

Op:rator terminates dilution

< 5241 Rod Cluster Control Assembly Ejection -

l 1.

Begmning of Life Full Initiation of rod' ejection 0.0 Power Power range high neutron Cux -

0.05 high setpoint reached Peak nuclear power occurs 0.14 Rods begin to fall into core 0.55 Peak fuel average temperature occurs 2.36

)-

i (01 OCT 1991)

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

t-

.l -

1 -

i f-i j

9.

REFERENCES

-l o

1

}

1.

Catawba Nuclear Station, Final Safety Analysis-Report, Docket l

4 Nos. 50 - 413/414.

T

+

l 2.

BAW-10172P-A, Mark-BW Mechanical. Design Report, Babcock &.Wilcox,-

~

{

Lynchburg, Virginia, December 19, 1989, i

3.

DPC-NE-200lP-A, Rev. 1, Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel, Duke Power Company, October 1990.

4.

BAW-10084A, Rev. 2, Program to Determine In-Reactor Performance f

of B&W Fuels - Cladding Creep Collapse, Babcock &'Wilcox,. October 1978.

l S.

BAW-10141P-A, Rev.-1, TACO 2 - Fuel Performance Analysis. Babcock i

& Wilcox, June 1983.

{

6.

DPC-NF-2010A, McGuire Nuclear Station / Catawba Nuclear Station

{.

Nuclear Physics Methodology for Reload-Design, Duke Power Company, June 1985.

3 4

7.

DPC-NE-2011P-A, Nuclear Design Methodology for Core Operating i

Limits of Westinghouse Reactors, Duke Power' Company, March 1990.

!=

8.

DPC-NE-2004P-A, McGuire and-Catawba ~ Nuclear Stations Core l

Thennal-Hydraulic Methodology using VIPRE-01,. Duke Power i

Company, December 1991.

v i

9.

BAW-10159P-A, BWCMV Correlation of Critical Heat Flux in-Mixing Vane Grid Fuel Assemblies, Babcock & Wilcox, July.1990.

1 i~

10, BAW-10173P-A, Mark-BW Reload Safety Analysis for Catawba'and l-McGuire, Babcock & Wilcox, Revision:2 February 20, 1991.

11.

DPC-NE-3000P, Duke Power Company, Thermal-Hydraulic Transient-Analysis Methodology, Revision 2, February-20,-1990.

12.

DPC-NE-3001-PA, ~ Duke Power Co:apany, Multidimensional Reactor-

[

Transients and Safety Analysis Physics Parameters Methodology, November 1991.

Ij' 13.

BAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba and-McGuire Units, Babcock & Wilcox, Revision 1, November 1990.

L

.14.

BAW-10168-A,.B&W Loss-of-Coolant Accident Evaluation Model:For-I.

Recirculating Steam Generator Plants', Babcock & Wilcox,.

i Lynchburgh, Virginia, January 1991.

15.

DPC-NE-1003A, Revision 1, McGuire Nuclear Station / Catawba Nuclear Station ^ Rod Swap Methodology ReportLfor Startup Physics Testing, December 1986.

i.

F U

l l

9-!

i-

. cu,; ; m ;a.-,

..,~:,.,,,-__,-

1

.i 9.

REFERENCES l.

Catawba Nuclear-Station, Final Safety Analysis Report, Docket Nos. 50 - 413/414, 2.

BAW-10172P-A, Mark-BW Mechanical Design Report, Babcock & Wilcox, Lynchburg, Virginia, December 19i.1989.

3.

DPC-NE-2001P-A, Rev. 1. Fuel Mechanical Reload Analysis Methodology for Mark-BW' Fuel, Duke Power Company,. October 1990.

4.

BAW-10084A,-Rev. 2, Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, Babcock & Wilcox. October 1 1978.

5.

BAW-10141P-A, Rev. 1, TACO 2 - Fuel Performance Analysis, Babcock

& Wilcox, June 1983.

6.

DPC-NF-2010A, McGuire Nuclear. Station / Catawba Nuclear Station Nuclear Physics Methodology for-Reload Design, Duke Power Company, June 1985.

~

7.

DPC-NE-2011P-A~, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, Duke Power Company, March 1990.

B.

DPC-NE-2004P-A, McGuire and Catawba Nuclear Stations Core.

. Thermal-Hydraulic Methodology using VIPRE-01, Duke Power Company, December 1991.

9.

BAW-10159P-A, BWCMV Correlation of Critical Heat' Flux in Mixing Vane Grid Fuel Assemblies,~ Babcock & Wilcox, July 1990, 10.

BAW-10173P-A, Mark-BW Reload Safety Analysis.for Catawba =and McGuire, Babcock & Wilcox, Revision 2, February 20, 1991, 11.

.DPC-NE-3000P, Duke Power Company, Thermal-Hydraulic Transient Analysis Methodology, Revision -2, February 20, 1990..

12.

DPC-NE-3001-PA, Duke Power Company, Multidimensional. Reactor Transients'and Safety Analysis Physics Parameters Methodology,

-Revision 2,_ November 1991.

13.

BAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba.and McGuire Units, Babcock & Wilcox, Revision 1, November 1990.

14.

BAW-10168-A, B&W Loss-of-Coolant-Accident Evaluation'Model For Recirculating Steam Generator' Plants,'Dabcock & Wilcox,-

Lynchburgh, Virginia, January 1991.

15.

DPC-NE-1003A, Revision 1, McGuire Nuclear Station / Catawba Nuclear Station Rod Swap Methodology Report for Startup' Physics-Testing, December 1986.

9-1 l

. -....... -..... -. - - - -. - -. ~. -

..... ~ - -

- -... ~

1:I J

I 1

i I

!=

ik i

16.

DPC-NE-3002-A,~McGuire Nuclear-Station / Catawba-Nuclear Station L

FSAR Chapter 15 System Transient Analysis Methodology, November j

1991.

d 17.

McGuire Nuclear Station Unit 1. Docket Number 50-369, Cycle 8=

Reload Submittal, Duke Power Company, June 26,1991.

-i j._

McGuire Nuclear Station Unit 2,~ Docket Number 50-370, Cycle 8 i

i 18.

j.

Reload Submittal, Duke Power Company, December 18,1991.

l 19.

Catawba Nuclear Station Unit 1, Docket Numbers 50-413-and 50-414, Cycle 7 Reload Submittal,-Duke Power Company, April,-13, 1992.

l-20.

Catawba-Nuclear Station Unit 1, Docket' Numbers 50-413 and 50-414, i-Cycle G Reload Submittal,_-Duke Power Company, January 9, 1991.

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-DPC-NE-3002-A,-McGuire Nuclear Station / Catawba Nuclear Station 16.

FSAR Chapter 15 System Transient Analysis Methodology, Revision 1 November 1991, '

I-17.

McGuire Nsclear Station Unit 1, Docket Number 50-369, Cycle 8 i

Reload Submittal, Duke Power. Company, June 26,19914 18, McGuire Nuclear Station Unit 2, Docket Number-50-370,-Cycle 8 i

Reload Submittal, Duke Power Company,-December 18,1991,.

i.

19.

Catawba Nuclear Station Unit 1, Docket Numbers 50-413 and 50-414, j

Cycle.7 Reload Submittal, Duke Power Company, April,- 13, 1992, 1

l 20.

Catawba Nuclear Station Unit 1. Docket Numbers 50-413 and 50-414, Cycle 6 Reload Submittal, Duke-Power Company, January 9, 1991.

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A Technical Justification i

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The Technical Specification (TS) and COLR changes as noted in Tables B-1 and 8-2 are identical to those previously submitted and approved for 2

j Catawba Unit 1 Cycle 7.

These changes reflect the transition from l.

Westinghouse to B&W supplied f uel and to Duke analysis methodology.

Recent cycles where changes were submitted and approved involving this j

transition are Catawba Unit 1 Cycles 6 & 7, McGuire Unit 1 Cycle B and McGuire Unit 2 Cycle 8.

The three exceptions to these changes, as i

denoted by an asterisk in Table 8-1, are new changes which have not been previously submitted.

i Procosed Revision to Technical Soecification 2.1.1 & Ficure 2.1-lh This proposed Technical Specification revision deletes Figure 2.1-lb and uses the current Figure 2.1-la to reflect use of the BWCMV CHF correlation and Duke Power Company's Statistical Core Design (SCD) j methodology with a 1.55 thermal design DNBR limit.

B Technical Justification i

These proposed revisions are the same as those approved for Catawba 1 l

Cycle 7 (Reference 9).

Procosed Revision to Technical Snecification Table 2.2-1 j

This proposed Technical Specification revision changes the K values for l

the overtemperature and overpower AT trip functions to reflect the use 1

of the BWCMV CHF correlation and Duke Power Company's Statistical Core Design (SCD) methodclogy with a 1.55 thermal design DNBR limit.

In addition, an axial imbalance penalty, f ( AI), is applied to the OPAT 2

reactor trip.

The power range neutron flux negative rate reactor trip l

is' deleted from the Reacter Protection System.

i Technical Justification l

These proposed revisions are the same as those approved for Catawba 1 l

Cycle 7 (Reference 9).

Procosed Rev_igj;n t Technical Scecification 3/4.2.1 f

This proposed revision provides Axial Flux Dxfference (AFD) limits l

consistent with Duke Power Company methodology.

Technical Justification i.

The _ proposed revisions are the same as the Axial Flux Dif ference (AFD) limit changes in the approved submittal for Catawba 1 Cycle 6 (Reference 10).

2 Procosed Revision to Technical Scecification 3/4.2.2 i

Specification 3/4.2.2 was revised to reflect the power peaking surveillance method described in DPC-NE-20llPA.

These revisions'are summarized as follows:

1.

The statement of the LCO was revised to reflect new nomenclature for the heat flux hot channel f actor [(Fo(X,Y,Z)] required by the

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f methodology in DPC-NE-20llPA and.used throughout the Reload j

-Report.

Also, as discussed in the McGuire 2 Cycle 8 reload submittal -(Ref erence 8), separato K(Z) curves are provided fors the Mark-BW and OFA fuel types.

1 2.

Action a in the_ current specification has been replaced by Actions a,:b,

-and c in the new specification.

The thermal power l

reduction required when Fi(X,Y,Z) ~ exceeds its limit are the same as the current requirement, as is-the reduction required in the OPAT trip setpoints. Action b is a new requirement,: and11s provided to limit the allowable AFD when Fi(X,Y,Z) exceeds its limit.

This reduces the possibility of operating the core in j

excess of the Fo(X,Y,Z) limit when a margin calculation = (discussed j

in item 7 below) indicates negative operational margin exists.

3.

.There is no change to SR 4.2.2.1.

4.

SR 4.2.2.2-addresses obtaining an incore flux map and the requirements-based on the results of the measurement..The l

reference to RAOC operation has been deleted, since RAOC j

operation is unique to Westinghouse methodology.

5.

There is no change to SR 4.2.2.2.a.

6.

SR 4,2.2.2.b in the current surveillance has been deleted.

The j

allowances for measurement uncertainty'and manufacturing tolerances have been-included in the limit { F$ 0<, Y, Z) ] and therefore the measured peak F5 0<,Y, Z) is not increased by these factors.

1 7.

SR 4.2.2.2.c in the current surveillance has been deleted.

No j.

simple determination is made of only whether or not the limit has i

been exceeded.

Instead, the amount by which the 4.2.2.2 measured i

value is above or below the limit is qualified as' detailed in item 10, below.

8.

SR 4.2.2.2.d (current surveillance) specifies the' frequency for 2

i measuring the core power distribution. This is done'by part.b in 4

the new surveillance.

Part'b.3 has been added to this I

surveillance, requiring an Fo(X,Y,Z) measurement when the excore -

quadrant power tilt ratio is normalized using incore detector 4

measurements.

This' ensures that_the impactiof any core tilt on

[

Fo(X,Y,Z) will be determined and reflected in the margin calculations of_part c.

j 9.

SR 4.2.2.2.e has been replaced by SR 4.2.2.2.d in the new j

surveillance.

The intent of:these requirements is similar in that projections of the measurements are made to determine at i

what point peaking would exceed allowable limits if the current trend continues.

In the new surveillance, an incore flux map is obtained and a determination is made as to whether the measured i

Fo(,',Z) will exceed _the: allowable peaking at 31 Effective Full XY

- Power Days (EFPD) beyond the most ecent measurement.

If the-extrapolated Fo(X,Y,Z) measurement exceeds the -allowable Fo(X,Y, Z) -

limit,-then-either the surveillance interval to the next power-a~

-distribution; map is_ decreased based on the available margin,:or the Fo(X,Y,Z) measurement is increased by 2% and the margin calculation of 4.2.2.2.c repeated.

This surveillance helps-i

-ensure that peaking will not exceed allowable limits prior to the i

next 31 EFPD measurement interval, f

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-The new SR 4.2.2.2.c replaces 4.2.2.2.f in the current j

surveillance.

The purpose of part c.1 is to perform margin j

calculations based on the measured peaks.

With the-now compared is the design pea)n at steady-state conditions, methodology, the limit. ([F (X,Y,Z))) to'which the measurement is i

increased

?

by a factor that represents the maximum amount that the power at i.

the given assembly location and axial elevation can increase above the design value before the measured value may become limiting.

Margins to both the LOCA peaking limit-(operational i

margin) and the centerline fuel melt limit (RPS margin) are calculated.

The operational margin forms the basis for 4'

restricting the AFD limits in part c.2, and the RPS margin forms l

the basis for reducing the OTAT trip setpoint in part c.3.

11.

SR 4.2.2.2.c.2 (new) replaces SR 4.2.2.2.f.2 in the current surveillance.

The reduced AFD limits determined in part-c.2 are 4

based on the amount of negative operational margin resulting from the margin calculation of part c l.

The negative and positive AFD limits are reduced 1% for each percent change in margin.

The AFD must be controlled to these new limits to reduce F,(X,Y,Z),

j and to-ensure that peaking will be. limited-for continued power

}i

- 12.

SR 4.2.2,2.c.2.b (new) corresponds to SR 4.2.2.2.f.2.b (current operation.

3 j _

surveillance).

13.

Part 4.2.2.2.c.3 has been added to the surveillance.

This part-1 j

of the surveillance requires reducing the Ki value of the OTAT trip setpoint if the RPS margin is negative.

This requirement i

i ensures that centerline fuel melt protection exists when core i

peaking may be greater than'the design values.

1 i

14.

SR 4.2.2.2.f.2.c, which addresses Base Load operation, has been i

deleted from the new surveillance.

The power distribution j

methodology of DPC-NE-2011PA does not constrain core operation to a target AFD.

i 15.

SR 4.2.2.2.g has been replaced by SR 4.2.2.2.e in the new i

surveillance; there are no. substantive changes to this-surveillance.

e l-16.

SR 4.2.2.3 addresses Base Load Operation and has b'een deleted from the new surveillance.

?

17.

SR 4.2.2.4 addresses surveillance of peaking in Base Load operation and has-been deleted from the new surveillance.

1 I.

-18.

SR 4.2.2.5 has been replaced-by SR 4.2.2.3 in the new surveillance; there are no substantive changes to this

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surveillance.

j Technical Justification

[

These proposed revisions-are thejsame as those approved for:McGuire 2 Cycle 8-(Reference 8).

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-Pronosed Revision to' Technical soecification 3/4.2.3 j

i Specification 3/4.2.3 was revised to reflect the power peaking surveillance method. described in DPC-NE-20llPA.

These revisions are summarized as follows:

~

1 1

1.

The statement of the.LCO was revised to reflect new nomenclature i

for the nuclear-enthalpy rise hot channel f actor (FL(X,Y)] and j

related parameters required by the methodology of -DPC-NE-2011PA r and used throughout the Reload Report..

4 i

i 2.

Those requirements of Actions a and 'b in the current.

i specification relating to the Reactor Coolant System flow rate have been incorporated in Specification 3.2.5.- The Actions.have 4-been revised to include the reduction of allowable thermal power when FU(X,Y) exceeds the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The factor-(RRR),

l

. by which the power level is decreased per percent F (X,Y) is j

above the limit, is defined in the COLR.

The inverse of-this 1

i factor is the fractional increase in the MAPS allowed when thermal power is decreased by 1% RTP.

When a power level.

decrease is required because _F,(X,Y)- has exceeded its ~ limit, then Action b requires restoration of Fu(X,Y). to within its limit within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,: or a reduction in the high flux trip setpoint.

The amount of reduction-of the high flux trip setpoint is governed by the same-factor (RRH) that determines the thermal power level reduction.

This-maintains core protection and an j

operability margin at the reduced power level similar to that at i

r rated thermal power.

3.

Action b.3 was replaced by Action d.

The portions of-the Action.

i requirements related to Reactor Coolant: System flow rate have

}:

been incorporated in Specification 3.2.5, and do-not appear in Action d of the new specification.

I j

4.

Action item c has been added and requires a1 reduction in the OTAT K trip setpoint by an amount equivalent to TRH for each 1%-

i i

Fa(X,Y) exceeds its limit within-72-hours of: initially being-k'

.outside the limit.

This' action ensures that the one protection-margin-is maintained at the. reduced power level for DNB:related transients not covered by the reduction in the Power. Range Neutron Flux-High Trip Setpoint.

5.

There is.no change to SR 4.2.3.1.

i 6.

SR 4.2.3.2 formerly covered only surveillance frequency.

It has been expanded es detailed below to reflect the power peaking i

surveillance' method described in DPC"NE-2011PA and the. format'of i

the revised SR 4.2.2.2.

Part a addresses. obtaining an incore flux map.

j 7.

ful 4.2.3.2.b (new)i replaces-the current'4.2.3.2;and addresses the j

frequency ' for' confirming that F (X,Y) is within its limit.

In -

i addition to performing the surveillance at least once per 31 EFPD, the revised surveillance requires' measurement of the.

j.

peaking 1 factor whenever the excore quadrant power tilt' ratio'is normalized using.incore detector measurements.

This ensures that-l the impact of any core-tilt on' Fm(X,Y) will be determined and j

reflected in the_ margin calculation. 'This is comparable to the new: SR 4.2.2.2.b in the F (X,Y,Z) specification.

.The surveillance o

requiring a surveillance'to be performed prior t'o operation above

-75%-of RATED THERMAL. POWER at the beginning of each fuel cycle _

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has been replaced by the requirement identical-to SR 4.2.2.2.b.2 in _ the F,(X,Y,Z) specification.

This surveillance ensures that l-the plant is at equilibrium conditions prior to a measurement, and also has a provision similar to the requirement it replaced j

stating that during power escalation at.the beginning of each cycle, THERMAL PONER may be increased until a power level for 4-extended operation has been achieved and a power distribution map j

obtained.

1 The purpose of part col is to 8.

SR 4.2.3.2.c has been added.

i perform margin calculations based on the measured radial peak.

The-limit [Fh(X,Y))" to which the measurement is compared is 3

I based on the allowable design.MARP limit, increased by a factor that represents the maximum amount that.the power-at the given j

asseably location can increase above the design value before the j

measured value may become-limiting.

Part c.2 uses the amount of margin determined by this procedure to form the basis for the l

amount of power level reduction and the reduction in the high flux and OTAT X trip setpoints required in the ACTION statements i

for the specification.

This is comparable to the new SR 4.2.2.2.c on F (X,Y,Z).

o 1

9.

SR 4.2.3.2.d has been added.

This surveillance requires projections of the measurements to be made to determine at what point Fa(X,Y) would. exceed the allowable-limit if the current trend continues.

In part d.1 a penalty is applied to.FL(X,Y).if the trend indicates that the-measured peak would exceed the limiting peak within the 31 EFPD surveillance period, and the margin calculations are repeated.

This provides additional i

margin, or-a buffer, to help-ensure that the peak will not exceed the limit prior to next 31.EFPD measurement interval.

In part d.2, the measurement is obtained and the margin calculations are i

repeated so that appropriate actions can be taken before zero margin is reached. This surveillance ensures the core is monitored at a frequency that considers conditions when measured j

peaks are-underpredicted.

This is comparable to the new SR

4. 2. 2. 2.d on Fo (X, Y, Z),

i t

10.

HSR 4.2.3.3, 4.2.3.4, and 4.2.3.5 in.the current specification j-address measurement of Reactor Coolant System: flow rate.

These requirements have been incorporated in Specification 3.2.5, and have been deleted from the revised requirements for SR 4.2.2.

Technical Justification i

These proposed revisions'are the same as those approved for McGuire 2 Cycle 8'(Reference 8).

l Procosed Revision to Technical Soecification'3/4.2.4 This proposed revision is intended to provide Quadrant' Power Tilt Ratio limits consistent with-Duke Power Company methodology.

l 1

Technical Justification

-)

i The proposed revisions are the same as the Quadrant Power Tilt Ratio

~ limit changes in the approved ~ submittal for Catawba-1 Cycle:6 j

(Reference 10).

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i Pronosed Revision to Technical Boecification 3/4.2.5 This proposed revision is intended to provide DND parameter limits consistent with Duke Power Company methodology.

Technical Justification The proposed revisions are the same as the DNB parameter limit changes l

in the approved submittals for Catawba 1 Cycle 6 (Reference 10) which revised the DND parameter limits consistent with Duke Power Company methodology and Catawba 1 Cycle 7 (Reference 9) which corrected a j

typographical error.

Pronored Revision to Technical Soecification Table 3.3.;

This change is to delete the reactor trip on power. range neutron flya negative rate from the Reactor' Protection System.

Technical Justification This proposed revision is the same as that approved for Catawba l' Cycle 7 (Reference 9).

Procesed Revision to Technical Snecification Table 3.3-2 The reactor trip on power range neutron flux negative rate is deleted.

Heutron detector response time exemption is added to OPAT trip.

Technical Juntification These proposed revisions are the same as those approved for Catawba 1 Cycle 7 (Reference 9).

Pronosed Revision to Technical Boecification Table 4.3-1 This change is to delete the reactor trip on power range neutron flux negative rate from the Reactor Protection. System.-

Technical Jut

'&ation This proposed rision is the same'as that approved for Catawba 1 Cycle 7 (Reference 9).

Procosed Revision to Technical sngelfication Table 3.3-4 This proposed revision changes the low steam line' pressure setpoint for safety injection and main steam line isolation-from 725 psig-to 775 psig... The allowable-value for this trip function is changed from 6941 psig to=744 psig, maintaining the same 31 psig allowance for rack; i

= uncertainties, and the-lead-lag controller-for. steam line pressure-low is. deleted.

Technical Justification These proposed revisions.are the same.as those approved for Catawba.1:

Cycle 7 (Reference 9)..

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pr inosed nevision to Technical roccification_ Table 3.3-9 Two response times are modified in this proposed change, the feedwater isolation response time is changed from 7 seconds to 12 seconds and the steam line isolation time is changed from 7 seconds to 10 seconds.

j Igchnical Justification i

i These proposed revisions are the same as those approved for Catawba-1 f

Cycle 7 (Reference 9).

Prononed Revision to Technical Eoecification 3.3.3.12 & 4.3.3.12.2 It is proposed that the reactor makeup water pump flowrate limit for Mode 5 be changed to 70 gpm in Technical specification 3.3.3.12(a)(2),

j

3. 3. 3.12 (b) (2) & 4.3.3.12.2(b)

Technical Justification j

1 Catawba is equipped with a Doron Dilution Mitigation System which-

}

serves to detect uncontrolled dilution events-in Modes 3 - 6 of plant j

operation.

The BDMS uses two source range detectors to monitor the suberitical multiplication of-the reactor core.- An alarm setpoint is e

j continually calculated as four times the_ lowest count rate, including i

compensation for background and the statistical variation in the count i

rate.

Once the alarm setpoint is exceeded, each train of the BDMS will

]

automatically shut off both reactor makeup water pumps, align the suction of the charging pumps to highly borated water from the Refueling Water storage Tank,'and isolate flow to the charging pumps from the Volume Control Tank.

Since these functions are automatically i

actuated by the BDMS, no operator action is necessary to terminate the 1

dilution event and recover the shutdown margin.

In the event one or more trains of the-BDMS is inoperable, the reactor makeup water pump l

flowrate limits ensure that the operator has sufficient time to recognize.and terminate a boron dilution event prior to the loss of i

shutdown margin during each appropriate mode of plant operation.

Each j

cycle, a bounding ratio of initial to critical boron concentration is.

established from the reload _ design.. This ratio is used to calculate the maximum reactor makeup water. pump flowrate which satisfies the 3

operator action time _ acceptance criteria of the Standard Review Plan, 1

The limits on reactor makeup water pump flowrates when the Boron Dilution Mitigation System (DDMS) is inoperable are. verified each cycis to ensure the safety analysis assumptions for these' parameters remain 4

valid. When the calculated reactor makeup water flowrate is found.to be less than the existing.flowrate limits, the-flowrate limits must be reduced such that-the operator action time acceptance criteria can be met.

These cycle-specific parameter limits are verified using the NRC 1

approved methodology provided in the attachment to a Duke Power letter to the U. S. Nuclear Regulatory Commission,- "... Supplementary Information Relative to Topical Report BAW-10173: Boron. Dilution Analysis", dated May 15, 1991 (Reference. ' h and Catawba FSAR

-(Reference 11) Section 15.4.6..

It is woposed that the reactor makeup' F

water flowrate limit for Mode SJbe reduced to-70 gpm.--This new flowrate limit is required to satisfy the operator-action time acceptance criteria in the Standard Review Plan.-

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Procosed Revision to Technical Boecification 3.4.2.1 & 3.4.2.2 l

This modification changes the tolerances on the pressurizer safety valve lif t setpoint from 11% to 43%,-2% in all modeo of operation.

Iechnical Justification f

These proposed revisions are the same as those approved-for Catawba 1 I

Cycle 7 (Reference 9).

l Procosed Revision'to Technical Boecification Table 3.6-2a & 3.6-2b l

This change clarifies the required maximum stroke time of the steam generator main feedwater to auxiliary feedwater nozzle isolation valves, auxiliary nozzle temper valves, steam generator feedwater containment isolation valves, steam generator feedwater purge valves, main steam isolation valves, and main steam isolation bypass control valves.

The numerical value of the stroke time of these valves is changed to HA.

1 Technical Justification The justification for the change in' valve stroke-time as it relates to system thermal-hydraulic. response during a steam line break event was-presented for a change to Technical Specification Table 3.3-5 in the Catawba Unit 1 Cycle 7 reload submittal (Reference 9).

Although'these valves are included in Tables 3.6-2a and 3.6-2b, the list of containment isolation valves for Unit 1 and Unit ~2, these valves do not receive a containment isolation-signal.

As shown in Catawba FSAR Figure 7-2, Part 8 of 16 a containment pressure high signal, low pressurizer pressure signal, low steamline pressure signal or a safety injection signal will actuate feedwater isolation-in addition to and separate from a phase

'a' isolation.

Also, a containment high-high signal,1cne steamline pressure, or high steam pressure rate signal will actuate a steamline isloation in addition to and separate from a phase

'b' isolation.- The valves in the proposed change are " actuated by signal other than S, T, or P signal (main steam isolation, feedwater isolation, low RN pit-level....)' according to note 8 of Catawba FSAR Table 3-104. These valves perform a containment isolation function-only to the extent that credit for their operation might be taken in the dose analysis.

Since these valves receive no containment isolation signal, and credit for the operation _of these valves is not taken in the dose analysis, a maximum stroke time is'not applicable for these

-valves.

Procosed Revision to Technical: Soecification 4.7.114 The permissible'strokeitime'for-the main steam isolation valves is I

changed 1from 5 to 8 seconds.-

i' Technical Justification-si lThis proposed revision-is the same as that approved for Catawba 1 Cycle 7 (Reference-9)4 k

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Erononed l<evinion to Technical Snecification 6.9.1.9 Add NRC approved Topical DPC-!!E-1004A,* Nuclear Design Methodology Using CASMO-3/SIMU! ATE-3P' to list of analytica) methods used to determine the core operating limits.

Technical Justification This change is administrative in nature since it updates the reference list with a newly approved topical describing methodology used to determine core operating limits.

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l References 1

1.

BAW-10159P-A, BWCMV Correlation of critical Heat Flux in Mixing Vane Grid Fuel Assemblies Babcock & Wilcox, July 1990.

2.

DPC-NE-2011P-A, Duke Power Company, liuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, March, 1990.

3.

DAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units, Dabcock & Wilcox, May 1991.

4.

DPC-NE-3001P, Duke Power Company, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, Revision 1, November 1991.

5.

DPC-NE-2004P-A, Duke Power Company, McGuire and Catawba Nuclear---

Stations Core Thermal-Hydraulic Methodology using VIPRE-01, December 1991.

6.

WCAP-10988, Cobra-NC, Analysis for a Main Steamline Dreak in the Catawba Unit 1 Ice Condenser Containment, Westinghouse Nuclear Energy Systems, November-1985.

l 7.

McGuire Nuclear Station Unit 1, Docket Number 50-369, Cycle 8 Reload Submittal, Duke Power Company, June 26,1991..

8.

McGuire Nuclear Station Unit 2, Docket Number 50-370, Cycle 8 Reload Submittal, Duke Power Company December 18,1991.

9.

Catawba Nuclear Station Unit 1, Docket Numbers 50-413 and 50-414.-

Cycle 7 Reload Submittal, Duke Power Company, April, 13, 1992.-

10.

Catawba Nuclear Station Unit 1, Docket Numbers 50-413 and 50-414, Cycle 6 Reload Submittal, Duke Power Company, January 9, 1991.

11.

Catawba Nuclear Station, Final Safety Analysis Report, Docket Nos. 50-423/414.

I 12.

Duke letter to U. S Nuclear Regulatory Commission, McGuire Huclear Station Docket Numbers 50-369 and.-370 Catawba Nuclear Station Docket Numbers 50-413 and -414 Supplementary Information-Relative to Topical Report BAW-10173; Doron Dilution Analysis, Duke Power Company, May 15,1991.

13.

McGuire Nuclear Station, Final Safety Analysis Report,, Docket Nos. 50-369/370.

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No Significant Hazards Evaluation 10 CFR 50.92 states that a

proposed amendment involves no significant hazards consideration if operation in accordance with the amendment would not:

1)

Involve a

significant increase in the probability or consequences of an accident previously evaluated; or 2)

Create the possibility of a now or different kind of accident from any accident previously evaluated; or 3)

Involve a significant reduction in a margin of safety.

CHANGES WHICH ARE THE SAME AS THOSE MADE FOR CIC7 The changes to the Safety Limit and Power Distribution Technical Specifications presented in Section 8

of the Reload Report represent the application of previously approved methodology to i

Catawba Unit 2.

The changes to remove the power rango neutron flux j

negativo rate reactor trip, increase the low steam line pressuro

~

setpoint, increaso. feedwater-isolation response time, increase l

steam line isolation-response time, increaso pressurizer safety l

valve lift setpoint tolerance, remove steam line pressure dynamic compensation, increase pressurizer safety valvo lift setpoint tolerance, and increase main steam line isolation valve stroke time 4

I reflect tho use of Duke analysis, and have already been approved l

for Catawba Unit 1.

The changes described above include the deletion of references to specific units on individual Technical j

specification pages, and dolate pages which were previously for Unit 2 only.

The implementation of unit specific references became j

necessary due to the transition from Westinghouse to B&W supplied fuel and for the Cycle 7 Reload due to the transition to Duke analysis methodology.

The analysis which made the changes necessary in the Unit i reload submittal is generic, and-as described:in the technical justification,-is equally applicable to l

both McGuire and Catawba _ units.

Therefore, there is no new l

significant hazards consideration which will be _ raised by this amendment.

This determination is_in keeping with staff guidance j

which was published in the Federal Register (4BFR14864) to. assist.

j in determiningiwhether or not_ proposed amendments are likely to i

raise a significant hazards consideration._ This guidance cites as i

an example of an amendment not - likely to involve 1a significant -

hazards consideration "a purely administrative change to technical -

specif1 cations: for example,-a. change to achieve consistency..."

Sinbe these changes are considered administrativo, - no further analysis is required.

i CHANGES TO TS 3/4.6.3 The proposed changes to the valvo stroke times in Table 3.6-2a and-3.6.2b will _not 'significantly_

increase-the probability or

- i consequences of any previously evaluated accident.

The effects of d

. -_ ~

l the dolays in isolation timos on the various transients affected have boon analyzed and found to be acceptable.

Since thoso valvos do not receive a containment isolation signal, and no credit is taken for operation of those valves in the dose analysis for a containment isolation function, a maximum stroke timo does not apply for containment isolation.

The proposed changes will not significantly increase the possibility of a now accident not previously ovaluated.

Foodwater and main steam isolation are responses to ongoing transients, rather than initiators or procursors of transients.

No equipment or ccmponent reconfiguration will occur as a result of this chango.

The proposed changes will not significantly decrease any margin of safety.

The isolation times which are applicable to thoso valvos are speciflod in Tablo 3.3-5, Engincored Safety Features Responso Times.

The offects of the isolation of those valvos was ovaluated based on their ESF function, not a containment isolation function, and determined to be acceptable, thorofore thoro is no significant decrease in the margin of safety.

CHANGE TO TS 3.3.3.12.a.2 TS 3.3.3.12.a.2 is changed to reduce the allowablo Reactor Makeup Water Pump flow in Modo 5 from 75 gpm to 70 gpm.

In the event that the Boron Dilution Mitigation System (BDMS) is inoperable the Reactor Makeup Water Pump flowrates are limited to ensure that operator action times required to terminato a dilution-ovent _can be met.

The limits on reactor makeup water pump flowrates when the BDMS is inoperable are verified each cycle to ensure that the safety analysis assumptions for those paramotors remain valid.

When the calculated Reactor Makeup Water Pump flowrato is found to be less than the existing flowrato limits, the flowrato limit must be reduced so that the operator action-timo acceptance critoria of Standard Review Plan 15.4.6 can be mot.

Reducing the allowable Reactor Makeup Water Pump flow in Mode 5 does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The current TS flowrate does not allow enough time for the operator to terminate an uncontrolled dilution event when required operator responso l

timos are assumed.

The lower flowrate allows nooded operator l_

response timos and is therefore moro conservativo.-

Reducing the allowable Reactor Makeup Water Pump -flow in Mode 5 does not change the way that any plant equipment is-operated:or maintained, therefore it does not croato the possibility of a new or different accident.

Reducing the Allowable Reactor Makeup Water Pump Flow in Mode 5 will not involve a significant reduction in the margin of safety..

This flowrate is more conservativo, and ensures that. safety analysis assumptions regarding operator actions times in response

to the termination of an uncontrolled dilution ovent can be mot.

Changos to TS 6.9.1.9 The proposed change to TS 6.9.1.9 adds approved topical DPC-NE-1004A to the list of analytical methods used to determine core operating limits.

This change is administrativo, adding a topical report which has boon approved for use on Catawba to the list of Sinco analytical methods used to dotormino core operating limits.

this chango is administrative it has boon determined that no significant hazards are involved.

The proposed Technical Specification change has boon reviewod against the critoria of 10 CFR 51.22 for environmental i

considerations.

As shown above, the proposed change does not involve any significant hazards consideration, nor increaso the types and amounts of ef fluents that may be released of fsite, nor increase the individual or cumulativo occupational radiation exposures.

Based on this, the proposed Technical Specification change meets the critoria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requiremont for an Environmental Impact Statomont.

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