ML19351A674

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Cycle 5 Core Operating Limits Rept.
ML19351A674
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 08/25/1989
From:
SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML19351A566 List:
References
NUDOCS 8912200429
Download: ML19351A674 (37)
v • d • e


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EXAMPLE i tc SOUTH CAROLINA ELECTRIC & GAS COMPANY 1

VIRGIL C. SUMMER NUCLEAR STATION- i

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. -CYCLE 5.

7 u4-CORE 0PERATING LIMITS REPORT +

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B912200429 891211 PDR ADOCK 05000395 P- PNU

Core Operating Limits Report i for V. C. Sum 2er Cycle 5 i

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, 1.0 CORE OPERATING LIMITS REPQRT EXAMPLE \

This core Operating Limits Report (COLR) for V. C. Summer Station Cycle 5 has teen prepared in accordance with the requirements of Technical Specification 6 9.1.11 The listedTechnical below Specifications affected by this report are 3.1.1 3 Moderator Temperature Coefficient 3.1 3.5 Shutdown Rod Insertion Limit 3.1 3.6 Control Rod Insertion Limits 3 2.1. Axial Flux Difference j

3.2.2 Heat Flux Hot Channel Factor 1 323 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor

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1 1

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( Page 1 of 27 t

4 . - . -- - _ .- ---

core Operating Limits Roport

. ,.. . EXAMPLE For V. C. Summer Cycle 5  !

1 2.0 OPERATING LIMITS listedfollow.

which in Section 1.0 are presented in the subsectio NRC-approved Specification 6 9.1.11. methodologies specified in TechnicalThes 2.1 i Moderator Temeerature Coefficient (Specification 3 1.1 3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits ares '

i The BOL/ARO-MTC shall be less positive than the limits shown la Figure 1  !

The EOL/ARO/RTP-MTC shall be less negative than t^

-5x 104 Ak /k /'F.

2.1.2 ,

The MTC Surveillance limit ist The 300 pps/ARO/RTP-MTC should Le less negative than or equal to -4.1x104 Ak/k/'F. -

where:

BOL stands for Beginning of Cycle Life ARO stands for All Rods Out RTP stands for RATED THERMAL POWER EOL stands for End of Cycle Life )

)

Page 2 of 27

.. 1 EXAMPLE 1

.9 h UNACCEPTABLE

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g _7 OPERATION W

2 ACCEPTABLE 5 .6 OPERATION i

.5 g 1 i h.4 ,

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0 10 20 30 40 50 60 70 80 90 100

% OF RATED THERMAL POWER FIGURE 1 MODERATOR TEMPERATURE COEFFICIENT V5 POWER LEVEL Page 3 of 27

, , , ,_,_.%_.. ._,m_, m,,..,--4,..y -...e.e,.-,-_-,w....,,...rmo,....m..m, ,._m ,,.# , , - _ - . - . , ., . - - , , - _ _

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

. 70It 09tPating Limit 3 , ,, '

For V. C. Summer Cycle 5 Report  :'

+. '

4 t

. 2.2 Shutdown Rod Insertion Limits  !

(Specification 3 1 3.5)

The shutdown steps. rods shall be withdrawn to at least 228 l 23 Control Red Insertion Limits (Specification 3.1 3 6)

The

2. Control Bank Insertion Limits are specified by Figure v EXAMPLE l i

4 l

l Page 4 of 27 >

EXAMPLE 230 0.54, 22s _-

200  !

/ 1,194 BANK C 180

/ [

/

160

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140  ;

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t / /

120  !, [

- 0, tis /

100 [

I ' / BANKD 80

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60

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40

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20

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/ 0.048,0 0 .1 .2 .3 .4 .5 .6 .7 .8 .9 1 INSERTED FRACTION OF RATED THERMAL POWER FIGURE 2 ROD GROUP INSERTION LIMITS VER$US THERMAL POWER FOR THREE LOOP OPERATION Page 5 of 27

-^-:;

Coro Operating Limits Report  :

For V. C. Summor Cycle 5

. j

, 2.4 Axial Flux Difference

' (Specification 3.2.1) 2.4.1 The Axial operation Flux Difference (AFD) Limits for RAOC for Beginning-of-Cycle Life (BOC)

Middle-of-Cycle Life (MOL) and End-of-Cycle Life  !

(EOL) are shown in Figures 3 through 5.

respectively.

The cycle burnup ranges applicable to each limit are indicated in each of the figures.  ;

2.4.2 The Axial Flux Difference (AFD) target bands during base load operating for BOL. MOL and EOL ares BOL (0 - 4000 MWD /MTU)  : + or - SS about a measured target value .

MOL-(4000 - 10000 MWD /MTU) : + or - 55 about a i measured target value r I EOL (10000 - 18000 MWD /MTU): + or - 5% about a asasured target value 2.4 3 The minimum allowable N power level for base load ,

operation APL D. is 85% of RATED THERMAL POWER. i i EXAMPLE t

i i

Page 6 of 27

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t EXAMPLE 120 110 '

100 *15.100

+10 too l \ '

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90 I \

' e ,_ . UNACCEPTABLE [ \ UNACCEPTABLE 80 I 5'

! ACCEPTABLE k 70 I l

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$ 60  ! k g I \

! 50 I E -

.31, so + 24, 50 f 40 30 20 I 10 0-50 -40 30 -

20 10 0 10 20 30 40 50 AXIAL FLUX DIFFERENCE (% DELTA 4) I

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FIGURE 3 i

AXtAL FLUX DIFFERENCE LIMITS A5 A FUNCTION OF RATED THERMAL POWER FOR CYCLE SURNUP 0 4000 MWD /MTU  ;

Page 7 of 27

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

. EXAMPLE 120 -

I 110 100 O. W

+ 10.1oo

/ (t ,

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90 \

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j q. UNAcetPTAsts

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50 -40 30 20 10 0 10 20 30 40 50 AxlAL Flux DeFransNcs(% osLTA4 i

I FIGURE 4 i

AXlAL FLUX DIFFERENCE UMfT5 A5 A FUNCTION OF RATED THERMAL POWER FOR CYCLE BURNUP 400010000 MWD /MTU Page 8 of 27

s i

EXAMPLE \

  • i 120 $

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110 1

100 14,100 i io,ioo '

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j AXIAL PLux ptFFantNCE(% DELTA 4 FIGURE 5 AXtAL FLUX DIFFERENCE LIMITS A5 A FUNCTION OF RATED THERMAL POWER FOR CYCLE SURNUP 10000 MWD /MTU . EOL Page 9 of 27 j  %

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-.,,.y., ,,. --,w ,g.,..,. .,-w-,n. . . , ,,,-,-,----,,--n.w

~ _ _ _ . ~ . - .

- - - - - ~ ' ^ ^ ~ ~ ^ " ~ ~ ~ ~ ~ ~

'Core O

' 'perating Limits Report EXAMPLE ""~" 1 2.5 Heat l \

Flux Hot Channel Factor - TQ(Z) (Specification 3.2.2) i FQRTP FQ(Z) 1 - -

  • K(Z) for P > o,$ )'

P l

1 FQRTP F0(Z) 1

  • K(Z) for P 1 0.5 l

THERMAL POWER L, where P= l RATED THERMAL POWER .i' 1

2.5.1 FQRTP

2.05  !

2.5.2 K(Z) is provided in Figure 6 2.5 3 Elevation dependent W(t) values for RAOC operation {

-(

at 150. 4000 10000, and 16000 MWD /MTU are shown in  !

Figures 7 through 10, respectively. This i I

l- information core height is sufficient to determine W (z) versus in the range of 0 MWD /MTU to EOL burnup.

Three point interpolation of the data in Fi ures 7

-through 9 is sufficient to determine RAOC W a) versus core height-between a Cycle burnup of 0 to 1

4000 MWD /MTU. I For Cycle burnupa between 4000 l MWD /MTU and E0., burnup, W(s) vu jus core height may be obtained through three point interpolation of the I

data in Ficures 8 through 10.

'2.5.4 Elevation dependent W(s)BL values-for base load operation between 85 and 1005 of rated thermal power L with the iten 2.4.2 specified target band about a E seasured target value at 150, 8000, and 16000 MWD /MTU are shown in Figures 11 through 13, respectively.

This information is sufficient to )

determine W(z)BL versus core height for burnups in the range of 0 MWD /MTU to EOL burnup through the use of three point interpolation.

l Page 10 of 27 4

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l_ SO{pW CORE HEIGHT, FT. T i FUEL O ,

FUEL p .

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I FIGURE 6 K(2) . NORMAUZED Fo(a) A5 A FUNCTION OF CORE HEIGHT Page 11 of 27 l

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,,,,-,-----,,,,wr,,,e -----,-,---,-e-,.,-a,---, , - - ~ - . . , , - - , -,,.e -

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EXAMPLE 1.5 t a

a  !

1.45 a

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4 1.35 '

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1 0 1 2 3 4 5 6 7 8 9 10 11 12 >

Cont HEIGHT (FEET)

FIGURE 7 V. C. SUMMER RAOC W(Z) AT 150 MWD /MTU Page 12 of 27

i EXAMPLE

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DATA FOR FIGURE 7 ,

l V. C. SUMMER RAOC W(z) AT 150 MWD /MTU Core Heloht W

,M  !

0.0000 Core Heioht g 1.0000 0.1600 6.0800 1.1024 1.0000 0.3200 6.2400 1.1073 1.0000  !

0.4800 6 4000 1.1109 i 1.0000 i 0.6400 6.b600 1.1140 t 1.0000 {

0.8000 6 7200 1.1165 1.0000

6.8800 0.9600 1.0000 1.1183 L- 7.0400 1.1200 1.0000 1.1195 i 1.2800 7.2000 1.1200 I 1.0000 i' 1.4400 1.0000 7.3600 1.1198 L

1.6000 7.5200 1.1188 1.0000 1 1.7600 7.6800 1.1173 1.4904 7.8400 1.1150 1.9200 1.4646 2.0000 40000 1.1119 1.4379 8.1600 1.1083 2.2400 1.4107 8.3200 1.1042 2.4000 1.3833 2.5600 84800 1.0986 1.3559 8.6400 1.0932 2.7200 1.3284 l 4.8000 1.0921 2.8800 1.3016 8.9600 1.0964 3.0400. 1.2781

-s 3.2000 9.1200 1.1060 1.2627 3.3600 9.2000 1.1155 1.2546 9.4400 1.1247 3.5200 1.2454 9.6000 1.1340 3.6800 1.2M6 9.7600 1.1423 3.8400 1.2238 9.9200 1.1513 4.0000 1.2127 10.000 1.1529 4.1600. 1.2023 10.240 t.1757 4.3200 1.1931 10.400 1.0000 4.4400 1.1837 10.540 1.0000 4.6400 1.1734 10.720 1.0000 '

4.8000 1.1626 10.8F0 1.0000 4.9600 1.1516  :

11.040 1.0000 '

5.1200 1.1398 11.200 1.0000 '

5.2000 1.1271 11.360 1.0000 5.4400 1.1160

%.520 1.0000 5.6000 1.1077 11.600 1.0000 5.7600 1.1007 1_ 11.840 1.0000 5.9200 1.0981 12.000 1.0000 i

Page 13 of 27

.-.--...-.-y,,,7py..y w. .,.,-g,-wg----- - - - -,, y-,,v,+,,-%m~uewwmew-e - * - - -me--- N

EXAMPLE 1.5 i

i 1.45  !

l 1.4 t

1.35 A

1.3 a

i f 1.25  : ,'

6 a

,, A

  • 1.2 ,

e a 1.15 A m a" Y i T ,i 4 8

1.1 1.05 1 ,

0 1 2 3 4 5 6 7

  • 8 9 10 11 12 CORE HEIGHT (FEET)

FIGURE 8 V. C. SUMMER RAOC W(Z) AT 4000 MWD /MTU r.

Page 14 of 27

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

t EXAMPLE DATA FOR FIGURE 8 V. C. SUMMER RAOC W(z) AT 4000 MWD /M Core Heicht V,j(,(1) 00000 Core Heicht g 1.0000 O.1600 6 0800 1.0000 1.1278 '

0.3200 6.2400 1.0000 1.1375 0 4800 6 4000 1.0000 1.1457 0.6400 6.5600 1.0000 1.1531 0 8000 1.0000 6.7200 1.;598 O.9600 6.8800 1.0000 1.1655 1.1200 1.0000 7.0400 t.1703 1.2800 1.0000 7.2000 1,4400 1.1741 1.0000 7.3600 1.1769 1.6000 ' 1.0000 7.5200 1.t 787 1.7600 1.3300 7.6800 1.1796 1.9200 1.3110 7.8400 1.1794-2.0000 1.2914 8.0000 1.1781 i 2.2400 1.2712 8.1600 1.1760 2.4000 1.2510 8.3200 1.1731 2.5600 1.2308 8.4800 1.1679 2.7200 1.2095 8.6400 1.1636 .

2.8800 1.1906 8.8000 1.1666 3.0400 1.1784 8.9600 1.1745 3.2000 1.1734 9.1200 l 1.1838 3.3600 1.1717- 9.2800 1.1925 3.5200 1.1691 9.4400 1.2026 '

3.6800 1.1661 9 6000 1.2155 3 8400 9.7600 1.1634 1.2286 4.0000 1.1622 9.9200 1.2413 4.1600 10.080 1.1605 - 1.2541 4.3200 1.1580 10.240 1.2672 4.4800 10.400 1.1549 1.0000 4.6400 1.1512 10.560 1.0000 '

4.8000 1.1468 10.720 1.0000 4.9600 1.1416 10.880 1.0000 5.1200 11.040 1.1359 1.0000 5.2800- 11.200 1.1297 1.0000 5.4400 11.360 1.1221 1.0000 5.6000 1,1143 11.520 1.0000 5.7600 11.680 1.1129 1.0000 M.9200 11.840 1.1178 1.0000 12.000 1.0000 ,

Page 15 of 27

d

. EXAMPLE 1 1.5 --

1.45

~~

.l 4

1.4 1.35 .j L A

1.3

=

i.

L 6 a

{ 1.25 a a - ,.

1.2

  • _ ,e s

b dF

.ah .a 1.15

  • ,a' ,

a a

< a ,

%% A ,' ,

1.1 i ,

N 1.05 1

0 1 2 3 4 5 6 7 3 g 10 11 12 CORE HEIGHT (FEET)

FIGURE 9 V. C. SUMMER RAOC W(Z) AT 10000 MWD /MTU Page 16 of 27

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

DATA FOR FlGURE 9

(

f V. C. 5UMMER RAOC W(z) AT 10000 MWD /MTU  !

Core Heicht Mil  !

00000 1.0000 Core Heiaht W J

0.1600 1.0000 6.0800 1.1255 0.3200 1.0000 6.2400 1.1352 5

0 4800 1.0000 6 4000 1.1444 6.5600 $

0.6400 1.0000 1.1527 0 0000 1.0000 6.7200 1.1600 0 9600 6.8800 1.1665 1.0000 1.1200 7.0400 1.1720 1.0000 5 1.2000 7.2000 1.1765 1.0000 1.4400 7.3600 1.1800 1.0000 ,

1.6000 7.5200 1.1824 1.0000 1.7600 7.6800 1.1839 1.3052 -

1.9200 7.8400 1.1843 '

1.2812 2.0400 8.0000 1.1837 1.2564 2.2400 8.1600 1.1421 1.2319 2 4000 8.3200 1.1795 1.2000 2.5600 84800 1.1767 1.1851 2.7200 86400 1.1770 1.1610 2.8000 8.8000 1.1839 1.1394 3.0400 8.9600 1.1927 1.1278 3.2000 9.1200 1.1999 1.1237 3.3600 9.2800 1.2062 1.1229 3.5200 9.4400 1.2144 1.1220 3.6800 9.6000 1.2270 1.1208 3.8400 9.7600 1.2436 1.1197 40000 9.9200 1.2619 1.1186 4.1600 10 000 1.2796 1.1172

'4.3200 10.240 1.2978 '

1.1155 4.4000 10.400- 1.0000 1.1136 4.6400 10.560 1.0000 1.1113 4.8000 10.720 1.0000 1.1087 4.9600 10.880 1.0000

't.1056  !

5.1200 11.040 1.0000 1.1020 1 5.2000 11 200 1.0000 1.0978 5.4400 11.360 1.0000 1.0939 5.6000 11.520 1.0000 1.0933 5.7600 11.600 1.0000 1.1025 5.9200 11.840 1.0000 1.1149  !

12.000 1.0000 '

l l l

Page 17 of 27 l

l

1.5 ,-

EXAMPLE l I  !

1.45 I

1.4 1.35  !

1.3 - 1 . ,

f '

i

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a f 1.25 ,

. mu '

e  % -

i &, e 1

r 1.2 '

4, ""

a a -

I s

1.15 d '

=

i .

4,c 1.1 '

t 1.05 1

0 1 2 3 4 5 6 7  :

8 9 10 11 12 -

CORE HEIGHT (FEET)

FIGURE 10 V. C. $UMMER RAOC W(Z) AT 16000 MWD /MTU Page 18 of 27

  • g .

s DATA FOR FIGURE 10 EXAMPLE i' V. C. SUMMER RAOC W(z) AT 16000 MWD /MTU k Core Heicht W,(g}

Sore Heicht JW 00000 1.0000 {

0.1600 6.0000 1.2179 1.0000 0.3200 6.2400 1.2281 1.0000 ,

0.4800 6.4000 1.2371 1.0000 i

0. M 6.5600 1.2445 1.0000 0.8000 6.7200 1.2504 1.0000 O.9600 6.8800 1.2548 '

1.0000 1.1200 7.0400 1.2577 1.0000 1.2000 7.2000 1.2590 1.0000 1.4400 7.3600 1.2587 1.0000 1.6000 7.5200 1.2568 1.0000 1.7600 7.6800 1.2535 1.2672 .

7.8400 1.2486 1.9200 1.2484  ;

8.0000 1.2425 2.0000 1 2291 8.1600 1.2347 2.2400 1.2084 ,

8 3200 1.2261 2.4000 1.1898 8.4800 1.2221 2.5600 1.1702  !

8.6400 1.2222 2.7200 1.1487  !

8.0000 1.2201 2 8800 1.1301 .

8.9600 1.2173 3.0400 1.1234 9.1200 1.2170 3.2000 1.1240 9.2000 1.2227 3.3600 1.1245 I 9.4400 1.23a ?

' 3.5200 1.1283 -,

L 9.6000 1.2503 3.6800 1.1320 9.7600 1.2683 l 3.8400 1.1393 9.9200 '

4.0000 1.2866 1.1461 10.080 1.3047 4.1600 1.1514 10.240 1.3236 4.3200 1.1569 10.400 1.0000 4.4800 1.1613 L 10.560 1.0000 4.6400 1.1647 . .

10.720 1.0000 l- 4.8000 1.1670 4,9600 10.080 1.0000 1.1686 ,

11.040 1.0000 5.1200 1.1689 11.200 1.0000 5.2000 1.1477 11.360 1.0000 5.4400 1.1700 11.520 1.0000 5.6000 1.1792 11.600 1.0000 5.7600 1.1924 11.840 1.0000 5.9200 1.2061 12.000 1.0000 Page 19 of 27 l'

l I

r--w+- -.e ---orw-vr.v,,w--e- -o-*

EXAMPLE 1.15 l

1.14 1.13 t

1,12 1.11 1.1 )

I 1.09 h4 '

a 1.08  :,

f 6' 1.07

'a. g*

1.06 "a, a' a

'g 1.05 'g '

"4 6 f" 1.04 46 1.03 ,

i '

1.02 q

1.01 1

0 1 2 3 4 5 6 7 8 9 10 11 12 I CORE HEIGHT (FEET) i FIGURE 11 V. C. SUMMER CASELOAD W(Z) AT 150 MWD /MTU '

Page 20 of 27 9

DATA FOR FIGURE 11 V.C. SUMMER SASELOAD W(z) AT 150 MWD /MTU j

)

. Core Heloht g Core Heicht g 1

l 00000 1000n 6 20 1.0542 l 0.1m 1.0000 0.3200

$ ' '# 1.0554 .

1.0000 0** l 0.4800 1.0528 1.0000 6.5600 l i,ogit 0.6400 1.0000 6.7200 i,oggg -l 0.0000 1.0000 0.9600 68m 1,o477 l 1.0000  ;

7.0400 1.0465 1.1200 1.0000 7.2000 j o43g 1.2800 i.oooo 7 3600 1.0432 1.4400 1.0000 7.5200 3o44g ,

1.6000 1,o g '

7.6800 1.0466 1.7600 ioggy 7.4400 1ogg3 ,

1.9200 1.0985 80000 1osoo  ;

2.0000 1.0982  :

8.1600 i,ogiy 2.2400 1,ogy, 8.3200 1.0533 2.4000 3,ogy, l 8 4400 1.0550 2.5600 1.0968 86400 1.0568 2.7200 1.0961 8.8000 i,oggg 28800 1.0952 4.9600 1.0609 30400 3 o,41 I t1200 1.0630 3.2000 j,ogy, 28 1.0650 3 3W 1.0915 9.4400 1.0669 '

3.5200 3,ogoo 9.6000 1.0688 3 6400 1.0884 9.7600 1.0706 i- 3.8400 1.0868 9.9200 1.0723

. 4.0000 1.0852 to e j,o73, 4.1600 1.0836 10.240 1.0754 4.3200 1.0019 I 10.400 1.0000 4 4800 1.0001 1 1.0000 ,

4.6400 1.0742 10 720 1.0000 48000 ,opsg 10 m 1.0000 4.9400 i,o741 11.040 i,ooog 5.1200 i,oyjg 11.200 1,000o 5.2000 i,g '

1 1.0000 5.4400 1.0672 1.520 1,000n 56000 1.0651 11.600 j,ooon 5.7600 j. g 11.840 i.oooo 5.9200~ 1.0606 12.000 3.0000 Page 21 or 27 L

- ~ . - , , , , , , - .w. x- - , . ,.ve.,,--~~,,--m-, - - - - , - - , - , , - - , _ - - , - -.. .,___ _ _ _ , - _ .- _-- . _ , - - - ,

_,-,e--, , - ., ,

= .

~

f EXAMPLE 1.15  ;

1.14  ;

i 1.13 -c .

1.12 l 1.11 'd ; i a .i L

1.1 a

h 1.09 '

i i

a 1.08 <

' k' f 1.07 i

k d

'% pam 1.06 \. . /

'a f 1.05 'As#  ;

1.04 1.03 i 1.02 1.01 1

0 1 2 3 4 5 6 7 g 3 10 11 12 Coat HeGHT(Fest)

FIGURE 12 V. C. SUMMER BASELOAD W(Z) AT 8000 MWD /MTU Page 22 of 27

-wem e- 3- +----.g y y w- w .-y -y .e s.a.y < - ----e -- w-- - - - - - ---.-an- ---ww---.W + -e - +

DATA FOR FIGURE 12 V. C. SUMMER BASELOAD W(2) AT 8000 MWD /MTU Core Heioht gg 0.0000 1.m0 Core Heicht JW 0.1M 6 0000 1.0555 1.0000 '

O.3200 6.2400 1.0539 1.0000 0.4400 6.4000 1.0534 1.0@

OM 6.5600 1.0538 1.WO 0 0000 6.7200 1.0545 1.0000 0.9600 6 8800 1.0556 1.0000 7.0400 1.0575 1.1200 1.0000 7.2000 1.0595 {

1.2400 1.0000 7.3600 1.0612 1.4400 1.0000 7.5200 1.0625 1.6000 1.0000 7.6400 1.0637 1.7600 1.1139  !

7 8400 1.0648 1.9200 1.1121 }

8.0000 1.0654 2.0000 1.1100 8.1600 1.0666 J 2 2400 1.1075 8.3200 1.0674 2.4000 1.1048 8.4000 1.0600 2.5600 1.1018 3 8.6400 1.0685 2.7200 1.0985 8.8000 1.0647 2.4000 1.0951 4.9600 1.0647 3.0400 1.0914 9.1200 1.0687 3.2000 1.0873 9.2000 1.0649 3.3600 1.0833 9.4400 1.0699 3 5200 1.0406 ,

9.6000 1.0719 3 6400 1.0787 9.7600 1.0739 3.8400 1.0769  !

9.9200 1.0755

  • 4.0000 1.0749 10.000 1.0771 4.1600 1.0731 10.240 1.0785 4.3200 1.0716 10.400 1.0000 4.4800 1.0704  ;

10.560 1.0000 4.6400 1.0692 10.720 1.0000 40000 1.0643 10.880 1.0000 4.9600 1.0674 11.040 1.0000 5.1200 1.0663 11.200 1.0000 5.2800 1.0651 11.M0 1.0000 5.4400 1.06M l 11.520 1.0000 5.6000 1.0614 11.600 1.0000 5.7600 1.0600 11.840 1.0000 L 5.9200 1.0579 1' 12.000 1.0000 l

Page 23 of 27

i EXAMPLE \

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1.15 ,  !

1.14 6 -

1.'13 '

a 1.12 , i a

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.6.

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1.05 1.04 1.03 1.02 l 1.01 r

1 0 1 2 3 4 5 6 7 8 9 10 11 12 COM HEIGHT (FEET)

FIGURE 13  :

V. C. SUMMER 8ASELOAD W(Z) AT 16000 MWD /MTU Page 234 of 27

=. . .

DATA FOR FIGURE 13 EXAMPLE \

V. C. 5UMMER BASELOAD W(z) AT 16000 MW i

Core Heicht {

JW Core Heicht  !

00000 1.0000 JW  !

0.1600 1.0000 6 0800 1.0616 0.3200 1.0000 6.2400 =

1.0647 i' 04800 1.0000 6 4000 1.0678 i O 6400 1.0000 5.5600 1.0706 0.8000 1.0000 6.7200 1.0730 0.9600 1.0000 68800 i 1.0752 1.1200 1.0000 7.0400 1.0771 I t

1.2800 1.0000 7.2000 1.0787 1.4400 1.0000 7.3600 1.0000 t 1.6000 1.0000 7.5200 1.0811 1.7600 1.1488 7.6800 1.0420  :

1.9200 7.8400 j 1.1436 1.0426 2 0000 1.1340 80000 1.0830 2.2400 1.1321 8.1600 1.0432 '

2 4000 1.1259 8.3200 1.0833 2.5600 8 4800 1.0430 1.1193

( 2.7200 8.6400 1.1123 1.0834 2.8800 8.8000 1.0000 1.1050 3.0400 8.9600 i 1.0976 1.0932  ?

3.2000 9.1200 1.0976 1.0897 3.3600 9.2800 1.1019 1.0828 3.5200 9 4400 1.1060 1.0799 3.6800 9.6000 1.1098 1.0787 '

3.8400 9.7600 1.1133 1.0766 '

4.0000 9.9200 1.1165 1.0744 4.1600 10.000 1.1192 1.0722 4.3200 10.240 1.1217 1.06M 4 4800 10.400 1.0000 1.0675 4.6400 10.560 1.0000 1.0651 4.8000 10.720 1.0000 1.0626 i 4.9600 10.880 1.0000 1.0600 5.1200 11.040 1.0000 1.0574 5.2000 11.200 1.0000 1.0557 5.4400 11.360 1.0000 1.0557 5.6000 11.520 1.0000 1.0571

  • 5.7600 11.680 1.0000 1.0580 5.9200 11.440 1.0000 1.0591 12.000 1.0000 Page 25 of 27 O >

, - - - - - - a , - , ~ _ e--.,-,,-->~~-+,,,,--.w-.. ,,--.-----a --,

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l EXA MPLE

< Core vcOperating Licits Report  !

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' l 2.6 RCS Flow Rate (Specification 3.2 3) and Nuclear EnthalDv Rise Hot Channel u Fa i

1

,, FL FAHRTP e (1 + PFAH ' (1-P))

i THERMAL POWER where P=

RATED THERMAL POWER  :

i 2.6.1 FAHRTP= 1.56 i

2.6.2 PFAH = 0 3 2.6 3 The Acceptable Operation Region from the combination of Reactor Coolant System total flow and R is provided in Figuro 14.

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Page 26 of 27

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

    • EXAMPLE MEASUREMENT UNCERTAINTIES OF 2.1 % FOR FLOW ANO 4.0% FORINCORE MEASUREMENTOF saARE FW INCLUDEDIN THI$ FIGURE 38

. I i 1 -

ACCT #tAbit UNACCIPTAtt! '

OPERATION atGiON OPtRAtioNptcioN 36 l

34 Ib .

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5 30 E '

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.9 .95 1 1.05 1.1 R . F=4un.HII.o + 0.M1.0 P)]

NOTE 2 When operating in this ressen, the restracted power levois shot he consadored to be 100% of rated thermal power (RTP)for Technscal Speelhcation Pipure 2.1 1 FIGURE 14 RCS TOTAL FLOW RATE VERSUS R FOR THREE LOOP OPERATION Page 27 of 27 4

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Attachment 4 I i

Safety Evaluation i

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j Attachment 4.to Document Control Desk Letter >

1* L -December 11,1989 c

-Page 1 of 3

, SAFETY EVALVATION FOR CHANGES TO

,f TECHNICAL SPECIFICATIONS DELETING CERTAIN CYCLE-SPECIFIC PARAMETERS Description ofAmendment Request:

Generic Letter 88-16, dated October 4, 1988, was issued to encourage licens m to prepare changes to Technical Specifications related to cycle-snecific parameters. These Technical Specification changes will relocate cycle-specific parameter limits from Technical Specifications to the Core  ;

Operating Limits Report (COLR). Presently the parameter limits in the Virgil C. Summer Nuclear Station T @ ical Specifications are calculated using hRC-

-approved methodologies. These limits are evaluated for every relecd cycle.

and may be revised periodically as appropriate to reflect changes to cycle-specific variables. This is an administrative burden on both the NRC ard South Carolir.a Electric & Gas Company.

The generic letter provided guidance to allow relocation of certain cycle-dependent core operating limits from the Virgil C. Summer Nuclear Station 2 Technical Specifications. ,This.would allow changes to the. values =of core *

operating limits without prior approval (i.e., license amendment) by the NRC, 4 provided an. NRC-approved methodology for the parameter limit calculation is followed. Thus, future Virgil C. Summer Huclear Station core reloads and other revisions will require a safety review in accordance with the ,

requirements of 10 CFR 50.59 instead of a prior NRC submitte.1; Currently, each parameter limit proposed in the COLR utilizes the approved nathodologies identified in the revised Administrative Controls section of this lirense amendment request. Virgil C. Summer Nuclear Station will use these methodologies when performing core' reload designs and when any other revisions are made.

The proposed technical specification chances concern the relocation of several cycle-specific core operating limits for Virgil C. Summer Nuclear Station =from Technical Specifications to the COLR. A'new definition of the COLR will be added to the Technical Specifications. Additionally, certain individual Technical. Specifications.will be amended to note that cycle-specific parameter limits are contained in the COLR. A COLR para raph

will be added to the Administrative Controls Section [which will replace the Peaking Factor Limit Report]. The COLR will be required to be submitted to the NRC to allow continuea trending of the cycle-specific parameters. SCE&G intends for-this.to be a permanent change to the VCSNS Technical Specifications.

-The proposed changes will reference the COLR for specific parameters and will ensure that cycle-specific parameters are maintained with the limits of the ,

COLR. The cycle-specific parameter limits proposed for relocation to the COLR-as part of this license amendment request include:

(a 3.1.1.3 Moderator. Temperature Coefficient (b 3.1.3.5 Shutdown Rod Insertion Limit I (c 3.1.3.6 Control Rod Insertion Limits  ;

d) 3.2.1 Axial Flux Difference -

e) S2.2 Heat Flux. Hot Channel Factor f) 3.2.S RCS Flow Rate ax Nuclear Entha k. rise Hot Channel Factor i

2

fAttachment 4 to Document Control Desk Letter

    • - December 11, 1989 Page 2 of 3 If approved, this Technical Specification change will not affect the plant equipment at VCSNS. The revision is administrative in nature and will only impact the manner in which the cycle .Mecific limits are referenced. These '

limits, when relocated to the COLR, will be implemented and controlled per VCSNS programs and procedures.

The propo:ed changes are consistent with the requirements of 10 CFR 50.36 and the staff's proposed policy for improving Technical Specifications, delineated in SECY-86-10. " Recommendations for Improving TS." The policy allows process variables such as core operational limits to be controlled by specifying them numerically in the Technical Specifications or by specifying the method of calculating their numerical values if the staff finds that the correct limits will be followed in operation of the plant. The proposed revision references the NRC-approved calculation methodologies. The development of cycle-specific core operating limits will continue to be performed by the referenced methodologies which have been accepted by the NRC.

Virgil C. Summer Nuclear Station (VCSNS) is requesting the amendment because it will eliminate the administrative burden of processing Technical Specifications changes for each refueling cycle. Also, the proposed changes to.the Technical Specifications are considered to be improvements and are consistent with the-NRC stated policy for improving Technical Specifications (52 FR 3788 February 6, 1987).

Safety Evaluation:

-The current Technical Specification method of controlling reactor physics parameters to assure conformance to 10 CFR 50.36 (which requires the lowest functional levels acceptable for continued safe operation) is to specify the p values determined to be within the acceptance criteria using an NRC-approved I calculation methodology. As previously dircessed, the methods used by VCSNS l to calculate these-parameter limits have been reviewed and approved by the i NRC and are cont,istent with the applicable limits in the Final Safety l

Analysis Report (FSAR). If approved, the proposed amendment will not alter the input parameters or the methodologies for calculating these limits.

The removal of cycle dependent variables from the Technical Specifications has no impact upon plant operation or safety. No safety-related equipment, safety function, or plant operations will oe altered as a result of this proposed changc. Since the applicable FSAP limits will be maintained and the Technical Specifications will continue to require operation within the core operational limits calculated by these NRC-approved methodologies, this proposed change is administrative in nature and does not affect the purpose of the Technical Specifications involved. Appropriate actions to be taken by VCSNS'if limits are violated will also remain in the Technical Specifications.

This proposed change will control the cycle-specific parameters within the

~ acceptance criteria and assure cnformance to 10 CFR 50.36 by using the

. -approved methodology instead of specifying Technical Specification values.

Tne COLR will document the specific parameter limits m sulting from South Carolina Electric & Gas Company calculations, including mid-cycle or other

-revisions.to parameter values. Therefore, the proposed change is in conformance with the requirements of 10 CFR 50.36.

l

.c ,

Attachment 4 to DocumentLControl Desk Letter

    • . s December 11, 1989-Page 3 of 3 Any changes to the COLR will be made in accordance with the provisions of 10 CFR 50.59. From cycle to cycle, the COLR will be revised tuch that the appropriate core operating limits for the applicable cycle will apply.

Technical Specifications will not be changed.

To summarize, SCE&G will continue to calculate the cycle specific parameter limits using NRC-approved methodologies which are consistent with all applicable limits of the plant safety analysis that are addressed in the Final Safety Analysis Report. Since the Technical Specifications will continue to require operation within these limits, the proposed change is

- administrative in nature. SCE&G agrees with the NRC conclusion (Generic Letter 88-16) +, hat changes can be made to these cycle specific limits using the specified methodologies without affecting nuclear safety. Based on the above discussion, SCE&G is requesting this amendment wnich will' allow cycle dependent variables to be removed from the Technical Specifications and placed in the defined Core Operating Limits Report.

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b. o December 11, 1989 Page 1 of 4 4 -

NO SIGNIFICANT HAZARDS EVALUATION FOR CHANGES TO TECHNICAL SPECIFICATIONS DELETING CERTAIN CYCLE-SPECIFIC PARAMETERS Backaround

, Generic Letter 68-16, dated October 4, 1988, was issued to encourage licensees.to prepare changes to Technical Specifications related to cycle-specific parameters. These Technical Specification changes will relocate cycle-specific parameter limits from Technical Specifications to the Core -

OperatingLimitsReport(COLR). Presently the parameter limits in the Virgil C. Summer Nuclear Station Technical Specifications are calculated.using NRC-approved methodologies. These limits are evaluated for every reload cycle and may be revised periodically as appropriate to reflect changes to cycle-specific variables. This is an administrative burJen on both the NRC and South Carolina Electric & Gas Company.

'The generic letter provided guidance to allow relocation of certain cycle-dependent core operating limits from the Virgil C. Summer Nuclear Station Technical Specification. This.would allow changes to the values of core operating limits without prior approval (i.e., license amendment) by the NRC, provided an NRC-approved methodology for the parameter limit calculation is followed. Thus, futurn Virgil C. Summer Nuclear Station core reloads and other revisions will require a safety revi w in accordance with the requirements of 10 CFR 50.59 instead of a prior NRC submittal.

Currently, each parameter limit proposed in the COLR utilizes the approved methodologies identified in the revised Administrative Controls section of this license amendment request. Virgil C. Summer Nuclear Station will use these methodologies when performing core reload designs and when any other revisions are made.

Froposed Chance The proposed technical specification changes cancern the relocation of several cycle-specific core operating limits for Virgil C. Summer Nuclear Station from Technical Specifications to the COLR. A new definition of the COLR will be added to the Technical Specifications. Additionally. certain individual Technical Specifications will be amended to note that cycle-specific parameter limits are contained in the COLR. A COLR paragraph will be added to the Administrative Controls Section (whicn will replace the Peaking Factor Limit Report). The COLR will be required to be submitted to the NRC to allow continued trending of the cycle-specific parameters.

L L

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Attachment 5 to Document Control Desk Letter

'0 +1 -December 11, 1989 Page 2 of 4 t j _ .

The proposed changes will reference the COLR for specific parameters und will '

ensure that cycle-specific parameters are maintained with the limits of the '

COLR. . The cycle-specific parameter limits proposed for relocation to the COLR_as part of this license amendment request include:

3.1.1.3 Moderator Temperature Coefficient

, 3.1.3.5 Shutdown Rod Insertion Limit 3.1.3.5 Control Rod Insertion Limits 3.2.1 Axial Flux Differ (nce 3.2.2 Heat Flux Hot Channel Factor ,

3.2.3 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor.

The' proposed changes are consistent with the requirements of 10 CFR 50.36 and .

the staff's proposed policy for improving Technical Specifications, delineated in SECY-86-10. " Recommendations for Improving TS." The policy '

allows process ~ variables such as core operational limits to be controlled by specifying them numerically in the Technical Specifications or by specifying the method of calculating their numerical values if the staff finds that the correct limits will.be followed in operation of the plant. The proposed revision references the NRC-approved calculation methodologies. The i development of cycle-specific core operating limits will continue to be '

performed by the referenced methodologies which has been accepted by the NRC.

The proposed changes.to the Technical Specifications are also considered to "

be_ improvements and are consistent with the NRC stated policy for improving

. Technical Specifications (52 FR 3788. February 6, 1987).

1.

Safety Evaluation The current Technical Specification method of controlling reactor physics parameters to assure conformance to 10 CFR 50.36 (which requires the lowest q functional levels acceptable for continued safe operation) is to specify the  !

values determined to be within the acceptance criteria using an NRC-approved (calculationmethodology. As previously discussed, the methodologies for l

l- calculating these parameter limits have been reviewed and approved by the NRC and are consistent with the applicable limits in the Final Safety Analysis '

Report (FSAR).

-The removal of cycle dependent variables'from the Technical Specifications j u has no impact upon plant operation or safety. No safety-related equipment, safety function, or plant operations will be altered as a retult of this ,

!- proposed change. Since the applicable FSAR limits will be maintained and the )

l' ' Technical Specifications will continue to require operation within the core operational limits calculated by these NRC-approved methodologies, this proposed change is administrative in nature. Appropriate actions to be taken if limits are violated will also remain in the Technical Specifications.

4 This proposed change will control the cycle-specific parameters within the acceptance criteria and assure conformance to 10 CFR 50.36 by using the approved methodology instead of specifying Technical Specification values.

The.COLR will document the specific parameter limits resulting from South l.' Carolina Electric & Gas Company calculations, including mid-cycle or other l

6

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' Attachment 5 to Document Control Desk Letter i n ' December 11, 1989

  • l Page 3 of 4-revisions to parameter values. Therefore, the proposed change is in conformance with the requirements of 10 CFR 50.36. l Any changes to the COLR will be made in accordance with the provisions of 10 CFR 50.59. -From cycle to cycle, the COLR will be revised such that the I appropriate core operating limits for the applicable cycle will apply. l Technical Specifications will not be changed. j Determination of Sionificant Hazards Pursuant to 10 CFR 50.91 South Carolina Electric & Gas Company has determined that operation of the iacility in accordance with the proposed

' license amendment rec"est .. n:, . n 11ve any significan< Sazards considerr ~ 3 ae; W*i ; - 'er.htions 1 'O CFR SC he followinC discuss, .'u .5es how th; osed amendment satisfies v.ct ./ the three standrv af ' M '^ r ' ' '

  • ) TF. '

. i . *. ere involve a s1gnificant increase in the pr W .) i or ..nc" .Fe Acc lent previously evaluated.

. e . am Oyt,le-sprcific & ,aatIng limits from the Virgil C.

e umrrer K. . lear Sta 'n R/! oiu ' sc lfic t ons has no influence er apact:r N rrobahility or tn sequeces if any accident previously fve % t'd. The cycle-specific m c?ct4 tion limits, although not in lechnical S h it antiens, will ** fo nowed in the operation of the Virgil C. Summer Nuclear Statioe. The pro;NSe f wendwnt still requires exactly the same actions to be taken when or if limits are exceeded as is required by current Technical Specifications. The cycle specific limits within the COLR will be implemented and >

controlled per VCSNS programs and procedures. Each accident analysis I addressed'in the Virgil C. Summer Nuclear Station Final Safety Analysis Report (FSAR) will be examined with respect to changes in cycle-dependent parameters, which are obtained from application cf the NRC-approved reload design methodologies, to ensure that the transient evaluation of new reloads are bounded by previously accepted analyses.

This examination, which will be performed per requirements of 10 CFR 50.59, ensures that future reloads will not involve a significant increase in the probability or consequences of an accident previously l evaluated.

l- The proposed change does not create the possibility of a new or 2).

different kind of accident from any accident previously evaluated.

As stated earlier, the removal of the cycle specific variables has no influence or impact, nor does it contribute in an way to the probability or consequences of an accident. No safety-related equipment, safety function, or plant operations will be altered as a result of this proposed chano. The cycle specific variables are calculated using the Em " ' methods and submitted to the NRC to allow tne Staff tr d' W to

'nd the values of these limits. The Technical Specifi.at *!i r.yt nue to require operation within the required rom opt m rc m'ts .<. appropriate actions will be taken when or if 1imitt . s.

~

~

Attachment 5 to Document Control Desk 1.etter r . Co December 11, 1989 i

' *; Page 4-of 4- '

Therefore, the proposed amendment does not in any way create the possibility of a.new of different kind of accident from any' accident previously evaluated.

3) The proposed amendment does not result in a significant reduction in >

the margin of safety. .

^

The margin of safety is not affected by the removal of cycle-specific

, cora operating limits from the Technical Specifications. The margin

'of safety presently provided by current Technical Specifications remains unchanged._ Appropriate measures exist to control the values of these cycle-specific limits. The proposed amendment continues to '

require operation within the core limits as obtained from'the NRC-approved reload design methodologies and appropriate actions to be

, taken when or-if limits are violated remain unchanged..

The development of the limits for future reloads will continue to conform to those methods described in NRC-approved documentation. In addition, each future reload will involve a 10 CFR 50.59 safety review to assure that operation of the unit within the cycle specific limits will not involve a significant reduction in a margin of safety.

Therefore, the proposed changes are administrative in nature and do -

not-impact the operation of Virgil C. Summer Nuclear Station in a manner that involves a reduction in the-margin of safety, p

Conclusion-The Commission has provided guidance concerning the applicatio.1 of the standards for determining whether a significant hazards consideration exists.

lThis guidance (51 FR 7750) includes examples of the type of amr.ndments that ,

are considered not likely to involve significant hazards considerations. The change proposed is similar to the examples of administrative changes identified in 51 FR 7750. Additionally, the proposed change is consistent '

, with the NRC policy for_ improving technical specifications (52 FR 3788) and the proposed. change is consistent with 10 CFR 50.36 and 10 CFR 50.59.

In view of the preceding, South Carolina Electric & Gas Company has L determined that the proposed license amendment does not involve any significant hazards considerations. ,

4 W - -

_ _ _ _ _ . . . _ .- s

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