Text

Proposed Tech Specs,Removing Cycle Specific Parameter Limits & Resolving Unnecessary Burden of Processing Changes to Update Such Limits for Each Core Reload
	ML20247L728
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Site:	Summer
Issue date:	09/19/1989
From:	; SOUTH CAROLINA ELECTRIC & GAS CO.
To:	;
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ML20247L721	List: ML20247L718; ML20247L728;
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Marked-Up Technical Specification Pages s90p25o000 890919 ADOCK 05000396 FDR PDC p

o DEFINITIONS SECTION PAGE

1. 0 DEFINITIONS Y

1.1 ACTI0N...................................................

1-1 N

1. 2 ACTUATION LOGIC TEST.....................................

1-1

i 1.3 ANALOG' CHANNEL OPERATIONAL TEST..........................

1-1 1.4 AXIAL FLUX DIFFERENCE....................................

1-1

'f **

1. 5 CHANNEL CALIBRATION......................................

1-1

1. 6 CHANNEL CHECK............................................

1-1 1.7 CONTAIMENT INTEGRITY....................................

1-2

1. 8 CONTROLLED LEAKAGE.......................................

1-2

1. 9 CORE ALTERAT!0N..........................................

1-2 l

14 Q.10 00SE EQUIVALENT I-131....................................

1-2 t

t %c 1.11 E-AVERAGE DISINTEGRATION ENERGY..........................

1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME.................

1-3 k

1.13 FREQUENC1' N0TATION.......................................

1-3 J, g 1.14 GASE005 RADWASTE TREATMENT SYSTEM........................

1-3 44 1.15 IDENTIFIED LEAKAGE.......................................

1-3 1.16 MASTER RELAY TEST........................................

1-3 1

b 4 1.17 0FFSITE DOSE CALCULATION MANUAL (00CN)...................

1-4 f

1.18 OPERABLE - OPERA 8I LITY...................................

1-4 1.19 OPERATIONAL MODE - M00E..................................

1-4 iT 1.20 PHYSICS TESTS............................................

1-4 N

1.21 PRESSURE BOUNDARY LEAKAGE................................

1-4 k

1.22 PROCESS CONTROL PROGRAM (PCP)............................

1-4

) O 1.23 PURGE-PURGING............................................

1-4 1.24 QUADRANT POWER TILT RATI0................................

1-5 1.25 RATED THERMAL P0WER......................................

1-5 i

, ta 1.26 REACTOR TRIP SYSTEM RESPONSE TIME........................

1-5 i

1.27 REPORTA8LE EVENT.........................................

1-5

'4 1.28 SHUTDOWN MARGIN..........................................

1-5

$l 1.29 SLAVE RELAY TEST.........................................

1-5 1.30 $0uoIr! CATION...........................................

1-5 1.31 SOURCE CHECK.............................................

1-5 1.32 STAGGERED TEST 8AS15.....................................

1-6 1.32 THERMAL P0WER............................................

1-6 1.34 TRIP ACTUATING DEVICE OPERATIONAL TEST...................

1-6 1.35 UNIDENTIFIED LEAKAGE.....................................

1-6 1.36 VENTILATION EXHAUST TREATMENT SYSTEM.....................

1-6 1.37 YENTING..................................................

1-6 TABLE 1.1 OPERATIONAL M00ES...................................

1-7

^

TABLE 1.2 FREQUENCY N0TATION..................................

1-8 SU MER - UNIT 1 I

Amendment No.

ADMINISTRATIVE CONTROLS SECTION PAGE Review......................................................

6-9 Audits......................................................

6-10 Authority...................................................

6 10 Records.....................................................

6-11 6.5.3 TECHNICAL REVIEW AND CONTROL Activities..................................................

6-11 6.6. REPORTABLE EVENT ACTI0N.......................................

6-12 6.7 SAFETY LINIT VIOLATION.......................................

6-12 6.8 PROCEDURES AND PR0 GRAMS.................................*......

6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS Startup Report..............................................

6-14a Annual Report...............................................

6-15 Annual Radiological Environmental Operating Report..........

6-16 Sosiannual Radioactive Effluent Release Report..............

6-16 Monthly Operating Report....................................

6-18 (c*M, 0.e*M7taLW"'.fL Report..........................

6-18

)

6.9.2 SPECIAL REP 0RTS.............................................

6-18 j

6.10 RECORD RETENTION.............................................

6-18 6.11 RADIATION PROTECTION PR0 GRAM.................................

6-20 6'.12 HIGH RADIATisN AREA..........................................

6-20 SUPMER - UNIT 1 XIX Amendment No.

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall' exist when:

All penetrations required to be closed during accident conditions a.

are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.4.

b.

All equipment hatches are closed and sealed, Each air lock is in compliance with the requirements of Specification c.

3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and The sealing mechanism associated with each penetration (e.g., welds, e.

bellows or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of arty component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position, i

N DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

ABD L3 ear @

SUMMER - UNIT 1 1-2

t INSERT A-( ADD TO PAGE 1-2)

CORE OPERATING LIMITS REPORT 1.9a The CORE OPERATING LIMITS REPORT (COLR) is the unit specific document that provides core operatinq limits for the current operating reload cycle. The cycle specific core opptating limits shall be determined for each reload cycle in accordance with Spe ification 6.9.1.11.

Plant operation within these operating limits is ade'ressed in individual specifications.

4 O

..a

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall beg (* TIE Edebo spelfied in the Co2E OPEf47t& tsMtf5 RfAotT(coa). The, d Efurt,J/-Il APPLICABILITY. -

2 - MODES 1 and 2 only#

- MODES 1, 2 and 3 only#

ACTION:

f y

"I k t4c doh 80l.

With the MTC more positive than the9faitC' _' 1'_T above, a.

operation in MODES I and 2 may proceed provided:

Bot.

1.

Control rod withdrawal limits are established and maintained r - ~

sufficient to restor + the MTC to Inss positive than the att/

Q d colp.

D._ :_; $within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2.

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

3.

In lieu of any other report required by Specification 6.9.1, a Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

mf g b.

With the MTC more negative than the limit

-]be in V

HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"With K,ff greater than or equal to 1.0

See Special Test Exception 3.10.3 SumER - UNIT 1 3/4 1-4 Amendment No.

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be detemined to be within its limits during each fuel cycle as follows:

The MTC shall be measureu and compared to the BOL limitf ' ~_ _ ~

l a.

prior to initial operation above 5% of I

RATED TH L

,a r each fuel loading.

b.

The MTC shall be measured at any THERMAL POWER.and compared to

( ". _ _ z

__n_

_"_'"7(all rods withdrawn, RATED THERMAL POWER l

/ condition) within 7 uro after reaching tration of 300 ppa.

In the event this comparison indicates the MTC is more negative than C0.1 1-7.- ~. ' 3 the MTC shall be

{"-

remeasured, and c ed to the EOL MTC limit 2" _ __:- __ _

tk C.0 L.Rj s dudessed) at le once per 14 EFPD during the remainder o' th' l

fuel cyc e,

_n

_n YL 200 g n g g & & w &

n h L ^: ^ ^"^ ^ # ' '

_ f ^- y c og g, SulmER - UNIT 1 3/4 1-5 Amendment No.

9 t

, REACTIVITY CONTROL SYSTEMS

~

^

~ ~ ~ ~

3/4.1.3 MVABLE CONTROL ASSDeLIES GROUP HEIGHT LIMITING Cole ITION FOR OPERTION 3.1.3.1 All full length (shutdown and control) rods which are inserted in the core shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES-18 and 28 ACT!0N.

n

m i k a.

a.

With one or more full length rods inoperable due ta being immovable

'as a result of' excessive friction or mechanicalJinterferience or

+

known to be untrippable, determine that the SHUTD0l88 MARGIN require-ment of Specification 3.1.1.1'is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in

, HOT.STAl(18Y within 6' hours.

b.

With more than one full length rod misa11gned from the group stip $

counter demand position by more than i 12 steps (indicated position)-,

be.in HOT STAWBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Withmorethanonefull'iengthrodinoperableduetoarodcontrol c.

I urgent failure alarm or obvious electrical problem in the rod control l, g i

system for greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ..be in HDT STAW8Y within the fol-i lowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I oQ d.

With one full length rod inoperable'due to causes other than. addressed l

y by ACTION a., above,' or misaligned.from its group step counter demand l

y @D - g '

height by more than

12 steps (indicated position), POWER OPERATION ~

may continue provided that within one ho,ur either:

j Ig i 1.

The rod is restored to 0PERABLE status within'the above

{i alignment requirements, or

$ tt 2.

The remainder of the rods in the group with the inoperable rod are P

p ~i alignedtowithin*12stepsofthe.inoperablerodwithinonehourl while maintaining the rod sequence and insertion limits e6didg-s e

{ (E eneses44=en64stue; the THERMAL POWER level shall be restricted g VQ pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue.provided that:

See Special Test Exceptions 3.10.2 and 3.10.3.

SUMMER - UNIT 1 3/4 1-14 Amendment No.

REACTIVITY CONTROL SYSTEMS SHUTDOW R00 INSERTION LINIT LINITING CONDITION FOR OPERATION S.Etrd$h A. N h A I & b ert a 3.1.3.5 All shutdown rods shall be_"_ ^^_

38 w.,11,. Coff APPLICA81LITY: MODES 1* and 2*f.

OPEg in 4.irners y,

REMR.T(ceLR).

With a maximum of one shutdown rod

~

except for surveillance testing pursuant to Specification 4.1..., with n one hour either:

the roQ }y $ $ Uspr.cLN mh CeLR; a.

et.

b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

h ted IM.

k 'Ma'em M

_w' p1:>1-n S *GoI.R) r

~

SUWEILLANCE REQUIREMENTS M b'NYk N 4.1.3.5 Each. shutdown rod shall be determined to be W y

m da, coat.

With n 15 minutes prior to withdrawal of any rods in control banks a.

A, 8, C or D during an approach to reacter criticality, and b.

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

"See special Test Exceptions 3.10.2 and 3.10.3.

With K,9f greater than or equal ta 1.0 Sl8MER - UNIT 1 3/4 1-20

+

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION Thecontrolbanksshallbelimitedinphysicalinsertionas_.ga5}'rA p

3.1.3.6

~ tKi_ c RE 69@ Afsn Limits REPoR. T (CoLR)

MYIfled.

APPLICABILITY: MODES la and 2*#.

Ld. b b 4a. Limih ~6cuu.v JM Pm Jat Jhu. Lo.p

.,CTION:

A p

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4,1.3.1.2, either:

Restore the control banks to within the limits within two hours, or a.

b.

Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the "

". z, _. Win., &dta; gud

d2, vn. cb COLR;et c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

"See Special Test Exceptions 3.10.2 and 3.10.3

With K,ff greater than or equal to 1.0.

StMMER - UNIT 1 3/4 1-21

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0.2 0.3 0

0A OA 0.7 0.8 0.9 1.0 FULLY FRACTION OF TED THERMAL POWER INSERTED FIG E 3.11 ROD OUP INSERTION Li S VERSUS THERMAL POW THREE LOOP OPERA ON SU)MER - UNIT 1 3/4 1-22 i

d

b M"

l 3L40K PA s,E

_..,,_e,..,.....,_.,..____...

- -- --- _ n. - z,x_ u.n a..__--

- - 3 DELETE 3 LANK FWU.RE F2orn Tech SPGCp 4ND AT>D To 0.o L R IF APPEo0 AL foe Two Loop o9E.EATine i s carances In THE Fu. TARE.

SUMMER - UNIT 1 3/4 1-23

_y Af*5 '

sNt, Tbt. Cot 2E Oa) 3/4.2 POWER DISTRIBUTION LIMITS OPER/)T) 6-UmiT 5 RE9adT 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a.

the allowed erational space " " __

^--'"M

~

" - - - ^ ' ' ' ' '

for Relaxed Axial Offset Control (RAOC) operat< on, or cott b.

within the target band specified in the @ about the target flux difference during base load operation.

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *.

ACTION:

a.

For RA0C operation with ndicated AFD outside of the applicable limits specified in the c

R COLR 1.

Either restore the indicated AFD to within the @ specified limits within 15 minutes.-or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux - High Trip setpoints to less than or equal 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

For Base Load operation above APL with the indicated AFD outside of the applicable target band about the target flux differences:-

1.

Either restore the indicated AFD to within the PFLR specified target band within 15 minutes, or 2.

Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes, THERMAL PDWER shall not be increased above 50K of RATED THERMAL POWE c.

unless the indicated AFD is within the applicable RAOC limits.

"See Special Test Exception 3.10.2 ND

- APL is the minimum allowable power level for base load operation and will be provided in the L l _ 'r^-- i- _.. ^_,_. Oper Specification 6. 9.1.11.

-- _ z core OPEf8% UmiTS REPoET)

SUMMER - UNIT 1 3/4 2-1 Amendment No.

8

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F LIMITING CONDITION FOR OPERATION 3.2.2 F (z) shall be limited by the following relationsh_ips:_

q f

F (z) 1 [M eD] [K(z)] for P > 0.5 F RTP q

T

' gyp]

F (z)

[K(z)] for P $ 0.5 q

g p, THERMAL POWER M

RAltD THERMAL POWER /

end K(z) se the '--- on mA.I Fa(4

=

"--#"^"

' - - _ _ ^ ^,... _. : " for a

^

given core height 4eest6eme q u y,, & c o g,g,

A_P, PLICA 8ILITY: MODE 1.

l ACTION.

With F (Z) exceeding its limit:

q Reduce THERMAL POWER at least N for each 3 F (z) exceeds the a.

q Itait within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Seitpofnts within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 3 for each 3 F (z),

N exceeds the limit.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMM. POWER any then be increased provided F (z) is demon-strated through incore mapping to be within its limitq tJau.

Fa

= tL., F Ldt a.h RATED THERMAL NER (RTP) q 5 7 "' O J A I As. C o E E 0 9 E g g T : s Li m i r s REPORT' (CoLR);

g

=

SUP9tER - UNIT 1 3/4 2-4 Amendment No. 56 1

a n_

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not appifcable.

4.2.2.2 For RAOC operation, F (z) shall be evaluated to determine if F (z) is within its Itait by:

9 9

Using the movable incore detectors to obtain a power distribution a.

map at any THERMAL POWER greater than 5% of RATED THERNAL POWER.

b.

Increasing the measured F (z) component of the power distribution q

map by 35 to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied, Satisfying the following relationship:

c.

F$#

F (z) 1

" f*I for P > 0.5

__V g

Owl.7..L F (z) a RTP g

a,,,,

g F (z) < "W (z) x 0.5 for P 1 0.5

" " III Q

F,en where (2)isthemeasuredF(z)i ased by the allowances q

gyp for manufacturing tolerances and me hF the F limit, K(t) is 'c' _ ' ~~ _ _

3 P is the relat we j[e, R(r) roment uncertainty, is g, g (g),3,,

POWER, and W(1) is the cycle dependent function that acco norest operat o distrihu+(an tra,nsients encountered during geots 1 given in the __

per :+ecmcanon e.,N.1.11.

d.

Meesuring (1) according to the following schedule:

CME c&

LinriqREpoE T) 1.

Upon achieving equilibrium conditions after exceeding by 10%

or more of RATED THERMAL POWER, the THERMAL POWER at which F (x) was last determined.

or g

2.

At least once per 31 Effective Full Power Days, whichever occurs first.

"During power escalation at the beginning of each cycle, power level sa increased until a power level for extended operation has been achieved and power distribution map obtained.

SIP 9tER - UNIT 1 3/4 2-5 Amendment No i

+

POWER O!STRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) e.

With the maximum value of F (z)

K(z) over the core height (z) increasing since the previous determination of (z) either of the following actions shall be taken:

(1)

(z) shall be increased by 2% over that specified in Specification 4.2.2.2c. or (2) F (z) shall be sensured at least once per 7 Effective Full Power Days until tw successive maps indicate that the maximum value of (z)

K(z) over the core height (z) is not increasing.

f.

With the relationships specified in Specification 4.2.2.2c..above not being satisfied:

(1) Calculate the maximum percent over the core height (z) that F (z) exceeds its liett by the following expression:

q

1)

N 1) - 1, x 100 for P > 0.5 y

x K(z)

RTP (1) X W 1)". 1 x 100 for P < 0.5 QX K(1)j

,1-o.s SU K R - UNIT 1 3/4 2-6 Aaenament No. 35 d

e

POWER DISTRIBUTION LINITS SURVEILLANCE REQUIREMENTS (Continued)

(2) One of the following actions shall be taken:

(a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the applicable AFD limits by 1% AFD for each percent F (z) exceeds its limits as q

determined in Specification 4.2.2.2f.1).

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alars setpoints to these modified limits, or (b)

Compl F (z)y with the requirements of Specification 3.2.2 for q

exceeding its limit by the percent calculated above, or (c) Verify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation.

g.

The limits specified in Specifications 4.2.2.2c., 4.2.2.2e., and 4.2.2.2f. above are not appilcable in the following com plane regions:

1.

Lower core region from 0 to ISE, inclusive.

2.

Upper core region from 85 to 1005, int.tusive.

4.2.2.3 Base Load operation is permitted at powers above APLE if the following conditions are satisfied:

Prior to entering Base Load operation, meintain THERPEL POWER above a.

E APL and less than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain Base Load operation surveillance (AFD within applicable target band about the target flux diffemnce) during this time period. Base Load operation is then pagitted providip POWER is maintained between APLE and APL or between APL and 1005 (whichever is most limiting) and Fg surveillance is maintained pursuant to Specification 4.2.2.4.

APL is defined as,the sinfaim value of:

Ff aptst,

x K(r) x 1005

%(z)xW(z)gt over the core heigl.t (z) where:

(z) is the measured F (z) q increased by the allowances for annuf tolerances and C"2 ' r 1 ') P esasurement uncertainty. The F limit i q

M~"__

" " O W(z)gg s the cycle dependent function that accounts i

g n

'l U

8L AA8-w per Spectrication 6.7.1.11.

CofE OPEEMiG utm75

^

KEpot3 SUP91ER - UNIT 1 3/4 2-6a Amendeent No

=

NWER DISTRIBUTION LIMITS SURVEILLANCEREQUIREMENTS(Eentinued) 1

{

Z b.

During Base Lead operation, if the THERMAL POWER is de ND APL then the conditions of 4.2.2.3.a shall be satisfie ow re-entering Base Load operation.

4.2.2.4 During Base Load Operation F (2) shall be evaluated to det F (z) is within its Ifsit by:

q g

Using the movable incere detectors to obtain a power distrib a.

map at any THERMAL POWER above APLE b.

Increasing the measured F (z) component of the power distri q

map by 35 to account for manufacturing tolera Verify the requirements of Specification 3.2.2 are satisfied es.

Satisfying the following relationship:

c.

f p (1) 1 for P> APLE where:

9 (z) is the measured F (z). The F limit is q

q

~

PistherelativeTHERMkLPOWER.

CP W(z) is the cycle cependent function that accounta for 11eited E

distribution transients encountered durino normal operation 0 ; K(20 M given in the

g(g g

5 cat

.9.1.11.

__.. _. _ _ 1

__ j per.

cota. oneATsw& s.sers Rsw d.

Measuring (z) in conjunction with target flux difference determination according to the following schedule:

1.

Prior to entering SASE LDAD operation after satisfying Sec-previous 31 EFPD witgthe relative thermal i

maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to napping, and 2.

At least once per 31 iffective Full Power Days.

With the maximus value of e.

((z)

K(z)

(z) either of the following actions shall be taken ove of SUP9tER - UNIT 1 3/4 2-6b Amendeent No.

,y POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 1.

. (2) shall be increased by 2 percent over that specified in

' y. 2. 2. 4. c, or 2.

(z) shall be measured at least once per 7 Effective Full Power Days until 2 successive seps indicate that the maximus value of Fl(z)

K(z) over the core height (z) is not increasing.

f.

With the relationship specified in 4.2.2.4.c above not being satisfied, either of the following actions shall be taken:

1.

Place core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied, and remeasure (z),or 2.

Compl F (z)y with the requirements of Specification 3.2.2 for q

exceeding its limit by the maximum percent calculated over the core height (z) with the following expression:

RTP dx100forP>APLE F

I a _

s P

g.

The limits specified in 4.2.2.4.c, 4.2.2.4.e. and 4.2.2.4,f above are not applicable in the following core plane regions:

1.

Lower core region 0 to 15 percent, inclusive.

2.

Upper core region 85 to 300 percent, inclusive.

4.2.2.5 nihan F (z) is measured for reasons other than meeting the requirements q

of Specification 4.2.2.2 an overall measured F (z) shall be obtained from a q

power distribution sep and increased by 35 to account for manufacturing tolerances and further increased by SE to account for measurement uncertainty.

SW94ER - UNIT 1 3/4 2-6c Amendeant No 1

g

DELETE FI6u2E h) TECH sPEcp RELOCATE K(2)

To cote.

1.2

\\

3. N I N i

N N

\\

l l 'x l

JN

\\

's

\\

.7

\\

(

h

\\

.3

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4

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.3

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2 5 t:

N N

N

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o 0

1 2

3 4

5 7

8 9

10 11 12 o

cons H54HT,FT.

OF FUE UE URE 3.21 NO.RMALIZED Fg(s)

AFUNCTION OFCOR EIGHT k

M SUMER - UNIT 1 3/4 2-7 AmendmentNol

F (C

f c% (L% &OGG OP62ATsAllr LImt7S REPolT(Yoa) asnu & M w 2sar w og g c

POWER DISTRIBUTION LIMITS i

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation %iiisiii g

Where:

6H H

L p - RAIED THERMAL POWER Fh=MeasuredvaluesofFhobtainedbyusingthemovableincore c.

detectors to obtain a power distribution map. The measuref valuesofFhshallbeusedtocalculateRsincef,Z^::

l

'~ ']

includes measurement uncertainties of 2.1% for flow and 4% for

/

incoremeasurementofFh,and APPLICABILITY: MODE 1.

TAL Ad5 T M J/ed YW ACTION:

[

With the combination of RCS total flow rate and R outs ide the region of accept-ableoperation[-

k 1b, col & s l

a.

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

1.

Restore the combination of RCS total flow rate and R to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, verify through intore flux mapping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Identify and correct the cause of the out-of-limit condition prior c.

to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION items a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incere flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation

~

3 prior to exceeding the

[

following THERMAL POWER levels:

c^ '

I

?

00 SUMER - UNIT 1 3/4 2-8 AmendmentNo.ff,fp,

INSERT 3 ;(ADD TO PAGE 3/4 2-0)

+

D DI d.

Pg The Fg limit at RATED T eensar. POWER specified in the COLR

=

e. Pgg the Power Facter Multiplier specified in the COLR.

3

~,

1

~

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION.

(Continued) 1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated RCS total flow rate and R shall be_.

determined to be within the region of acceptable operation Ur

' " ");

l Prior to operation above 75% of RATED THERMAL POWER after each fu a.

loading, and

~

mL eoL%

b.

At least once per 31 Effective Full Power Days.

=

4.2.3.3 The indicated RCS total flow ra M^ T&t least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whe region of acesptable operation C' N.1 ;t I

verified to be within the

[

the most recently obtained value of R obtained per Spwification 4.2.3.2, is assumed to exist.

4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.5 The RCS total flow rate shall be determined by measurement at least once per 18 months.

SumER - UNIT 1 3/4 2-9 Amendment No. M.

I 1

TELETE num r20m Tcal sPEcp

~

REL.D LATE To cot 2.

MEASWL1rNP2dCERTAI t$ OF 2.1% FOR FLOW AND 4.i% $4.t.' CORE M UREMENTOF FWAn ARE INCLUDED THl$ FIGURE 38 M X3PTA UNACCEPT opeaanow a on openAno sow 36 s

F g

i g '*

~

r y

n I

I R

O (1.00.28.te m a

(t.se.as.es N '

ON 28 (i.ac.27.m

/

(1.00.27.50) 25 M

/

~

/

9

.95 1.05 1.1 e

avt.ssit.a.s.attmit soft: wheneeweenesiem me eennnes es==nesses shes sensafesa to l seemsuteneed pauperesTP)for 1

FIGURE 3.2-2 l

TOTAL FLOW RATE VS. R 00P OPERATION j

SLH9ER - UNIT 1 3/4 2-10 Amendment No. 4, $$

?..

6" Qm wmdim &

)

E nk y L.y (EOL)

REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIE (Continued) involved subtracting the incre ntal change in the MDC associated with a core condition of all rods inserted most positive MDC) to an all roc.s withdrawn condition and, a conversion fo the rate of change of moderator density with temperature at RATED THERMAL R conditions. This value of the M C was then transfomed into the limiting NTC value,~ :

--'T_

"r-[The'MTC value presents a conservative value (with correc-tions for burnup and soluble boron) at a core condition of 300 ppe equilibrium boron concentration and is obtained by making these corrections to the limiting Eo,L MTC valuef :.: _ :: ' _ T )4---

The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confim that the NTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be nada critical with the Reactor Coolant System average temperature less than 551'F.

This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4) the reactor pressure vessel is above-its minime RTET temperature.

3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is i

available during each sode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps,

3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from 0PERABLE diesel generators.

With the RCS average temperature above 200*F, a minimum of two boron in-jection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide the required SHUTDOWN SumER - UNIT 1 B 3/4 1-2 Amendment No.

f),75 g

~

3/4.2 POWER DISTRIBUTION LIMITS w

I BASES The specifications of this section provide assurance of fuel integrity Q

during Condition I (Normal Operation) and II (Incidents of Moderate Frequency)

(1) maintaining the calculated DN8li in the core at or above the events by:

design limit during normal operation and in short-ters transients, and (2) limiting the fission gas release mechanical properties to within assum,ed design criteria. fuel pellet temperature, and c In addition, limiting the peak ifnear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS g

i acceptance criteria limit of 2200*F is not exceeded.

y) Q-a The definitions of certain hot channel and peaking factors as used in I

these specifications are as follows:

@v0 F (z)

Heat Flux Hot Channel Factor, is defined as the maximum local ki O

-4 heat flux on the surface of a fuel rod at core elevation I divided g

by the average fuel rod heat flux, allowing for manufacturing A

talerances on fuel pellets and rods; Fh Nuclear Enthalpp Rise Hot Channel Factor, is defined as the ratio of

@ -6.g, the integral of linear power along the rod with the highest integrated power to the average rod power.

j 3/4.2.1 AXIAL FLUX O!FFERENCE b

its on AXIAL FLUX DIFFERDICE (AFD) assure that the F (z) upper bound q

envolt.pe o' times the normalized axial peaking factor is not exceeded during el r normal operation or in the event of xenon redistribution following power changes.

The Ifeits on AFD will be provided in the per Technical Specification 6.5.1.11 i

Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their i

respective insertion itetts and should be inserted near their normal position for staa differen@ce obtained under these conditions divided by the fraction of R

-state operation at high power levels.

THEIDEL POER is the target flux difference at RATED THElpmL POWER for the associated core burnup conditions. Target flux differences for other THElp%L POWER levels are obtained by multiplyt the RATED THEIDEL POWER value by the appropriate fractions 1 THDDEL level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

e

(

SumER - UNIT 1 8 3/4 2-1 Amendment No. If

^

POWER O!STRIBUTION LIMIT BASES AKIAL FLUX DIFFERENCE (Continued) g j

At power levels below APL, the Itaits on AFD are defined in the l

E consistant with the Relaxed Axial Offset Control (RA0C) operating proce ure and Ifeits. These lief ts were calculated in a manner such that expected

{

operational transients, e.g., load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of time allowed outside of the limits at reduced power levels will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prevent operation in E

the vicinity of the APL power level.

GOLR E

At power levels greater than APL, two modes o ration are pemissible:

(1) RAOC, the AFD timit of which are defined in the and (2) Base Load g

operation, which.is defined as the maintenance of t within PFLR specifica-E tions band about a target value. The RA0C operating procedure above APL is the same as that defined for operation below APLE. However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with F (z) less than its limiting value. To allow operation at the q

maximum permissible power level the Base Load operating procadure iets the indicated AFD to latively small target band (as specified in the and 0.0LP El power swings (APL

< power < APL or 2005 Rated Thermal Power, whichever is lower). For Base Load operation, it is expected that the plant will operate within the target band. Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would enange sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon redistribution impact from past operation on the Base Load operation, a 24-hour waiting period at a power level abeve APL and allowed by RA0C is necessary. During this time period load changes and rod action are restricted to that allowed by the Base Load procedure. After the waiting period extended lase Lead operation is pemissible.

The computer determines the one minuta average of each of the OPERABLE excore detector outputs and provides an alam message f amediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels am: (1) outside the allowed delta-I power operating space (for RA0C operation), or (2) outside the allowed delta-I target band (for Base Load operation). These alams are active when power is greater than: (1) 5GE of RATED THEIDEL POWER (for RAOC operation), or (2) APLE (for Base Load operation). Penalty deviation sinutes for Base Load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

SumER - UNIT 1 B 3/4 2-2 Amendment No

O POWER DISTRIBUTION LIMIT b<

BASES 90k

?/4.?.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE AND h

NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR L

Q The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak N.{s local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance 4

criteria limit.

b Each of these is measurable but will normally only be determined periodically g

as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is ey sufficient to insure that the limits are maintained provided:

Control rods in a single group move together with no individual rod a.

insertion differing by more than i 13 steps, indicated, from the group demand position.

b.

Controlrodgroupsaresequencedwithoverlappinggroupsasdescribed[

ir. Specification 3.1.3.6.

'g g *g The control' rod insertion limits of Specifications 3.1.3.5 and c.

3.1.3.6 are maintained.

N d.

The axial power distribution, expressed in terms of AXIAL FLUX g'

DIFFERENCE, is maintained within the limits.

i pm j

h will be maintained sithin its limits provided conditions a.

hrough t

4 g

d. above are maintained. AsnotedonT:.._^^QRCSflowrateand may l

be " traded off" against one another (i.e., a low measured RCS flow rate is h

g 3

acceptable i { " --' g is also low) to ensure that th calculatedDNBRfb' s

)

will not be below the design DNBR value. The' relaxation of F as a function / -

g

/'

of THERMAL POWER allows changas in the radial power shape for all permissible /h roc ?nsertion limits.

( Ny A

R, as calculated 3.2.3 and used in' :._.

^ ^ Q accounts for Fh 1ess than or equal to This value is used in the various accident analyses where Fh influence's parameters other than DNBR, e.g., peak clad temperature and thus is the maximum "as measured" value allowed.

i Margin is maintained between the safety analysis limit DNBR and the h]g g4 design limit DNBR. This margin is more than sufficient to offset any rod bow

)

penalty and transition core penalty. The remaining margin is available for I

plant, design flexibility.

b When an F0 easurement is i.aken, an allowance for both experimental error m

\\

and manufacturing tolerance must be made. An allowance of 5% is appropriate d%

for a full core map taken with the incore detector flux mapping system and a y

3% allowance is appropriate for manufacturing tolerance.

SUMMER - UNIT 1 B 3/4 2-3 Amendment No.

POWER DISTRIBUTION' LIMIT BASES MAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE IST CEANNEL FACTOR (Continueo)

The hot channel factor F (z) is measured periodically and increased by a cycle and height dependent power factor appropriate to either RAOC or Base Load operation, w(z) or W(z)BL' t provide assurance that the limit on the hot channel factor, F (z) is met. W(z) accounts for the effects of normal opera-q tion transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. W(z)gg accounts for the more restrictive operating limits allowed by Base Load operation which result in less severe transient values. The W(z) and W(z)BL functions described above for l

3 normal operation are providea in the@eek._ " ru

}

Specif4 cation 6.9.1.11.

_.. 77_) per l

CcRE OPERA 7tn& OMITS REPcRT(cot.R)

WhenRCSflowrateandFharemeasured,noadditionalallowancesare necessary prior to compariioiiwith the limits ok __ "% Measurement

]% Y gj o

errors of 2.1% for RCS total flow rate and 4% for Fh have been allowed for in determining the limits off- = " 7 l

b i

d The 12-hour periodic surveillance of indicated RCS flow is sufficient to p

detect only flow degradation which could lead to operation outside the l

acceptable region of operation sh g 7 1^Q l

3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribu-8 tion satisfies the design values used in the power capability analysis. Radial g-power distribution measurements are made during startup testing and periodically during power operation.

y The limit of 1.02, at which corrective action is required, provides DNS and linear heat generation rate protection with x y plane power tilts. A 1.imiting tilt of 1.025 can be tolerated before the margin for uncertainty in F

i g s depleted. The limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the maximum q

allowed power by 3 percent for each percent of tilt in excess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when.one excore detector is inoperable the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incure detector monitoring is done with a full incore flux map or twu sets of 4 symmetric thimbles. These locations are C-8, E-5, E-11 H-3, H-13, L-5, L-11, N-8.

SUMMER - UNIT 1 B 3/4 2-4 Amendment No. M,

POWER DISTRIBUTION LIMIT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RIS NOT CMf?MEL FACIOR (Continued) 3/4.2.5 DNB PARAMETERS' 1

The limits on the DN8 related parameters assure that each of the parameters l

are maintained within the normal steady state anvelope of operation assuscd in I

the transient ar.d accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimus DN8

_.- throughout each analyzed transient.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to c.nsure that the parameters are restored within their limits following load changes and other expected transient operation.

i y

W

k b~

Y M STRfBUTION LIMIT

^

~

BASES 1

HEAT FLUX HOT CHANNEL FACTOR a FL and NUCLEAR ENTHALPY RISE H NN L ont nued with the initial FSAR tions and have been ically demonstrated adequate to mainta sinimum DNBR of 1.30 througho analyzed transient.

The our periodic surveillance of these parameters throu rument reado s sufficient to ensure that the parameters are restored within 1

s following load changes and other expected transient operation.

4 i

kEPLAcE Utth e

.c

~ tL ca.o h s,,u L Q s f 1

~

b 3/4.4 REACTOR COOLANT SYSTEM'

.~

~

t BASES 3/4.4.1 REACTOR COOLANT LOOPS ND COOLANT CIRCULATION The plant is designed to erste with all reactor coolant loops in

' operation, and maintain DNBR1._ ^.Cduring all normal operations and anticipated transients.

In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STAND 8Y within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERA 8LE.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERA 8LE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, therefore, be within the capaht11ty of operator recognition and cont.rol.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 300*F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

SUlHER - UNIT 1 B 3/4.4-1 I

s*

ADMINISTRATIVE CONTROLS Type of container (e.g., LSA, Type A, Type B. Large Quantity), and e.

Solidification agent (e.g., cement, urea formaldehyde).

f.

The radioactive effluent release reports shall include unplanned releases from site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a qucrterly basis.

The radioactive effluent release reports shall include any changes to the Process Control Program (PCP) made during the reporting period.

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports cf operating statistics and shutdown experience, in-cluding documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating L iort within 90 days in which the change (s) was made effective.

In addition, a re;.c of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted as set forth in 6.5 above.

RADIALPEAKINGFACTQfLIMITREPORT 6.9.1.11 The limits, the W(z) Functi s for RAOC and Ba Load operation and the valu or APLND (as required) s 1 be established r each reload core and i emented prior to use.

RERAG The met ology used to generate W(z) functions fo DC and Base Load DTN Oper on and the value for APL shall be those pr cusly reviewed and N

app ed by the NRC.* If cha s to these methods re deemed necessary y

Id5Ef.T w.

de evaluated in accord e with 10 CFR 50.59 nd submitted to the for O

. view and approval prior '

their use if the c e is determined to volve n unreviewed safety que on or if such a cha e would require amen nt of previously submitted d ntation.

A report containin AFD limits, the W(

functions for RA0C Base Load operation and th alue for APLND (as r tred) shall be provi to the NRC document contr desk with copies to t regional administrate and the resident inspector wi n 30 days of their i mentation.

ND Any info ion needed to support z),W(z) and APL 1 be by request from th RC and need not be in dedinthikgreport.

P-10216 P A "Relaxati of Constant Axial Offse ontrol-F Surveillance cchnical Specification g

SumER - UNIT 1 6-1B Amendment No. M. O,

I.

INSERT C (ADD TO PAGE 6-18)

CORE OPERATING LIMITS REPOh r 6.9.1.11 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm a.

surveillance limit for Specification 3/4.1.1.3, b.

Shutdown Bank insertion Limit for Specification 3/4.1.3.5, Control Bank Insertion Limits for Specification 3/4.1.3.6, c.

d.

Axial Flux Difference Limits, target band, and APLND for Specification 3/4.2.1, Heat Flux Hot Channel Factor, FgRTP, K(Z), W(Z), APLND and W(Z)st or e.

f Specification 3/4.2.2, f.

Nuclear Enthalpy Rise Hot Channel Factor, FagnTP, and Power Factor Multiplier, PFas, limits for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a.

WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION M ETHODOLOGY", July 1985 (W Proprietary).

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit,3.1.3.6 - Control Bank insertion Limit, 3.2.1 - Axial Flux Difference,3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

b.

WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", June 1983 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (FQ Methodology forW(Z) surveillance requirements).)

c.

WCAP-10266-P-A, REV. 2, "THE 1981 VER510N OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987 (W Proprietary).

1 (Methodology forSpecification 3.2.2 - Heat Flux Hot Channel Factor).

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

-r to The CORE OPERATING UMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

F O

e

+e g

1-_

l l

1 Sample Core Operating Limits Report

l EXAMPLE SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION CYCLE 5 CORE OPERATING LIMITS REPORT August 25, 1989 l-4 e

1 Cora Operating Limits Report.

For V. C. Summer Cycle 5 l

EXAMPLE l

1.0 CORE OPERATING LIMITS REPORT i

l This Core Operating Limits Report (COLR) for V. C. Summer Station Cycle 5 has been prepared in accordance with the requirements of Technical Specification 6.9.1.11.

The Technical Specifications affected by this report are listed below:

3.1.1 3 Moderator Temperature Coefficient 3.1 3 5 Shutdown Rod Iasertion Limit 3.1.3.6 Control Rod Insertion Limits 3.2.1 Axial Flux Difference 3.2.2 Heat Flux Hot Channel Factor 3.2.3 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor Page 1 of 27

__--_-------------]

Coro Opsrating Limits Report EXAMPLE For'V. C. Summer Cycle 5 e

2.0 OPERATING LIMITS 1

The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the subsections which follow.

These limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.11.

2.1 Moderator Temperature Coefficient (Specification 3.1.1 3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:

The BOL/ARO-MTC shall be less positive than the limits shown in Figure 1.

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

-5x 104 Ak /k /*F.

2.1.2 The MTC Surveillance limit is:

The 300 pps/ARO/RTP-MTC should be less negative than or equal to -4.1x10-4 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 l

Page 2 of 27

J EXAMPLE l

1

.9 N.

8 UNACCEPTABLE OPERATION y

,7 ACCEPTABLE go

.6 OPERATION 2

E

.5 1

3

\\

2

.4 i

\\

.2

(

2

.1

\\

e 0

10 20 30 40 50 60 70 80 90 100

% OF RATED THERMAL POWER FIGURE 1 MODERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL Page 3 or 27

FT Core Operating Limits Report For V. C. Summsr Cycle 5 2.2 Shutdown Rod Insertion Limits (Specification 3 1 3.5)

The shutdown rods shall be withdrawn to at least 228 steps.

23 Control Rod Insertion Limits (Specification 3.1 3.6)

The Control Bank Insertion Limits are specified by Figure 2.

I EXAMPLE Page 4 of 27

b EXAMPLE 230

! 0.54,228 200 1,194 -

BANK C 180 j

/

l l

160

/

140 E

/,

/

120

/

~3

- 0,118 O

[

$100

'/

o c

BANK D 80

/

60

/

40

'/

20

/

0.048,0 l

0

.1

.2

.3

.4

.5

.6

.7

.8

.9 1

FULLY FRACTION OF RATED THERMAL POWER FIGURE 2 - ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER FOR THREE LOOP OPERATION Page 5 of 27

Core Oparating Limits Report For V. C. Summnr cycle 5 l

2.4 Axial Flux Difference (Specification 3.2.1) i 2.4.1 The Axial Flux Difference (AFD) Limits for RAOC L

operation 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 are:

BOL (0 - 4000 MWD /MTU)

+ or - 5% about a measured target value MOL (4000 - 10000 MWD /MTU)

.t + or - 5% about a measured target value EOL (10000 - 18000 MWD /MTU) : + or - 5% about a measured target value 2.4 3 The minimum allowable power level for base load operation, APLND, is 85% of RATED THERMAL POWER.

EXAMPLE Page 6 of 27

EXAMPLE 120 110 100 15,100

+ 10,100 90 UNACCEPTABLE [

k UNACCEPTABLE w

/

1

[

ACCEPTABLE h

70 f

}

\\

/

(

j

/

\\

o 50 31,50

+ 24,50

{

40 30 20 10 0

50

-40

-30

-20

-10 0

10 20 30 40 50 AXtAL FLUX DIFFERENCE (% DELTA-1)

FIGURE 3 AXtAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR CYCLE BURNUP 0 - 4000 MWD /MTU Page 7 of 27

EXAMPLE

~

120 110 100

300

+ 10,100

\\

90 g

UNACCEPTABLE

[

k UNACCEPTABLE

\\

80 I

ACCEPTABLE h

70 E

\\

g

/

\\

A 60 g

I

\\

50

\\

~

g

-25,50

+ 24,50 40 30 20 10 0

-50

-40

-30

-20

-10 0

10 20 30 40 50 l

AX1AL FLUX DIFFERENCE (% DELTA-l)

FIGURE 4 AXlAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR CYCLE BURNUP 4000-10000 MWD /MTU Page 8 of 27 t

o EXAMPLE 120 j

110 100

-14,100

+ 10,100 90 UNACCEPTABLE

[

k UNACCEPTABLE

(

)'

E 80 E.

/

\\

ACCEPTABLE h

70 w

(

\\

1 l

/

(

j

/

\\

50 h

-30,50

+ 24,50 f

40 30 20 10 0

-50

-40

-30

-20

-10 0

10 20 30 40 50 AXtAL FLUX DIFFERENCE (% DELTA l)

FIGURE 5 AXlAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR CYCLE BURNUP 10000 MWD /MTU - EOL Page 9 of 27

EXAMPLE M ' W @ e"l'

2.5 Heat Flux Hot Channel Factor - FQ(Z) (Specification 3 2.2)

RTP pg FQ(Z) 5.

K(Z) for P > 0.5 P

BTP pg F (Z) 1

K(Z) for P 10.5 Q

THERMAL POWER where P=

RATED THERMAL POWER 2.5.1 FQRTP= 2.45 2.5.2 K(Z) is provided in Figure 6 2.5.3 Elevation dependent W(z) values for RAOC operation at 150, 4000, 10000, and 16000 MWD /MTU are shown in Figures 7 through 10, respectively.

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

Three point interpolation of the data in Figures 7 through 9 is sufficient to determine RAOC W(z) versus core height between a Cycle burnup of 0 to 4000 MWD /MTU.

For Cycle burnups between 4000 MWD /MTU and EOL burnup, W(z) versus core height may be obtained through three point interpolation of the data in Figures 8 through 10.

2.5.4 Elevation dependent W(z)BL values for base load operation between 85 and 100% of rated thermal power with the item 2.4.2 specified target band about a i

measured target value at 150, 8000, and 16000

)

MWD /MTU are shown in Figures 11 through 13, I

respectively.

This information is sufficient to

{

determine-W(z)BL versus core height for burnups in 1

the range of 0 MWD /MTU to EOL burnup through the use of three point ir.terpolation.

I 1

1 I

Page 10 of 27 l

l t

EXAMPLE i

1.2 1.1 1

.9

.8 6

.7 El

.6 Q=

.5 3*

.4

.3

,7 CORE HEIGHT (to KJ 0.0 1.0 60 1.0 12.0 0 925 0

I I

O 1

2 3

4 5

6 7

8 9

10 11 12 80gp CORE HEIGHT, FT.

TP OM p

FUEL FUEL 4

l FIGURE 6 K(z)- NORMALIZED Fo(z) AS A FUNCTION OF CORE HEIGHT Page 11 of 27 w_

EXAMPLE

~

1.5 b

L 1.45 s

1.4 a

1.35 A

1.3 a

l

^

E 1.25

^

3 a

A a

1.2 a

L a

^

1.15 a

a A

L i

6 M 4

,a 1.1 g.

1.05 1

0 1

2 3

4 5

6 7

8 9

10 11 12 CORE HEIGHT (FEET)

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

EXAMPLE DATA FOR FIGURE 7 V. C. 5UMMER RAOC W(z) AT 150 MWD /MTU Core Heicht W(z)

Core Heicht W(z) 0.0000 1.0000 6.0800 1.1024 0.1600 1.0000 6.2400 1.1073 0.3200 1.0000 6.4000 1.1109 0.4800 1.0000 6.5600 1.1140 0.6400 1.0000 6.7200 1.1165 0.8000 1.0000 6.8800 1.1183

0. % 00 1.0000 7.0400 1.1195 1.1200 1.0000 7.2000 1.1200 1.2800 1.0000 7.3600 1.1198 1.4400 1.0000 7.5200 1.1188 1.6000 1.0000 7.6800 1.1173 1.7600 1.4904 7.8400 1.1150 1.9200 1.4646 8.0000 1.1119 2.0800 1.4379 8.1600 1.1083 2.2400 1.4107 8.3200 1.1042 2.4000 1.3833 8.4800 1.0986 2.5600 1.3559 8.6400 1.0932 2.7200 1.3284 8.8000 1.0921 2.8800 1.3016

8. % 00 1.0968 3.0400 1.2781 9.1200 1.1060 3.2000 1.2627 9.2800 1.1155 3.3600 1.2546 9.4400 1.1247 3.5200 1.2454 9.6000 1.1340 3.6800 1.2346 9.7600 1.1423

~ 3.8400 1.2238 9.9200 1.1513 4.0000 1.2127 10.080 1.1629 4.1600 1.2023 10.240 1.1757 4.3200 1.1931 10.400 1.0000 4.4800 1.1837 10.560 1.0000 4.6400 1.1734 10.720 1.0000 4.8000 1.1626 10.880 1.0000 4.9600 1.1516 11.040 1.0000 5.1200 1.1398 11.200 1.0000 5.2800 1.1271 11.360 1.0000 5.4400 1.1160 11.520 1.0000 5.6000 1.1077 11.680 1.0000 5.7600 1.1007 11.840 1.0000 5.9200 1.0981 12.000 1.0000 1

Page 13 of 27 i

o

I' EXAMPLE c

~

v.

1.5 1.45 1.4 1.35 A

1.3 a

a a

f 1.25 4

5 A

6 a

1.2 g-a a

'd4 1.15 a

,'s a

'A A

3

'd 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 Page 14 of 27

1 EXAMPLE DATA FOR FIGURE 8 V. C. SUMMER RAOC W(z) AT 4000 MWD /MTU Core Heicht W(z)

Core Heicht W(z) 0.0000 1.0000

{

6.0800 1.1278 0.1600 1.0000 6.2400 1.1375 0.3200 1.0000 6.4000 1.1457 0.4800 1 0000

)

6.5600 1.1531 0.6400 1.0000

(

6.7200 1.1598 O.8000 1.0000 6.8800 0.9600 1.0000 1.1655 7.0400 1.1703 1.1200 1.0000 7.2000 1.17# ~.

1.2800 1.0000 7.3600 1.1/69 1.4400 1.0000 7.5200 1.5787 1.6000 1.0000 7.6800 1 1796 1.7600 1.3300 7.8400 i.1794 1.9200 1.3110 8.0000

".1781 2.0800 1.2914 8.1600 1.1760 2.2400 1.2712 8.3200 1.1731 2.4000 1.2510 8.4800 1.1679 2.5E00 1.2308 8.6400 1.1636 2.7200 1.2095 8.8000 1.1666 2.8800 1.1901 8.9600 1.1745 3.0400 1.1784 9.1200 1.1838 3.2000 1.1734 9.2800 1.1925 3.3600 1.1717 9.4400 1.2026 3.5200 1.1691 9.6000 1.2155 3.6800 1.1661 9.7600 1.2286 3.8400 1.1638 9.9200 1.2413 4.0000 1.1622 10.080 1.2541 4.1600

1. <05 0.240 1.2672 4.3200 1.1580 10.400 1.0000 4.4800 1.1549 10.560 1.0000 4.6400 1.1512 10.720 1.0000 4.8000 1.1468 10.880 4.9600 1.1416 1.0000 11.040 5.1200 1.1359 1.0000 11.200 5.2800 1.1297 1.0000 11.360 5.4400 1.1221 1.0000 11.520 5.6000 1.1143 1.0000 11.680 5.7600 1.1129 1.0000 11.840 1.0000 5.9200 1.1178 12.000 1.0000 Page 15 of 27 l

\\

y l

l

.__-___--__-____A

EXAMPLE 1.5 1.45 1.4 1.35 6

1.3 a

h E 1.25 3

a 1.2

6'

,a' 5

.ams

^

6 1.15

^

a

?

t 1.1

^

Hs 1.05 1

0 1

2 3

4 5

6 7

8 9

10 11 12 CORE HEIGHT (FEET)

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

EXAMPLE DATA FOR FIGURE 9 V. C. SUMMER RAOC W(z) AT 10000 MWD /MTU Core Heicht W(z)

Core Heicht W(r) 0.0000 1.0000 6.0800 1.1255 0.1600 1.0000 6.2400 1.1352 0.3200 1.0000 6.4000 1.1444 0.4800 1.0000 6.5600 1.1527 0.6400 1.0000 6.7200 1.1600 i

0.8000 1.0000 6.8800 1.1665 0.9600 1.0000 7.0400 1.1720 1.1200 1.0000 7.2000 1.1765 1.2800 1.0000 7.3600 1.1800 1.4400 1.0000 7.5200 1.1824 1.6000 1.0000 7.6800 1.1839 1.7600 1.3052 7.8400 1.1843 1.9200 1.2812 8.0000 1.1837 2.0800 1.2566 8.1600 1.1821 2.2400 1.2319 8.3200 1.1795 2.4000 1.2080 8.4800 1.1767 2.5600 1.1851 8.6400 1.1770 2.7200 1.1610 8.8000 1.1839 2.8800 1.1394 8.9600 1.1927 3.0400 1.1278 9.1200 1.1999 3.2000 1.1237 9.2800 1.2062 3.3600 1.1229 9.4400 1.2144 3.5200 1.1220 9.6000 1.2270 i

3.6800 1.1208 9.7600 1.2436 3.8400 1.1197 9.9200 1.2619 4.0000 1.1186 10.080 1.2796 4.1600 1.1172 10.240 1.2978 4.3200 1.1155 10.400 1.0000 4.4800 1.1136 10.560 1.0000 4.6400 1.1113 10.720 10000 4.8000 1.1087 10.880 1.0000 4.9600 1.1056 11.040 1.0000 5.1200 1.1020 11.200 1.0000 5.2800 1.0978 11.360 1.0000 5.4400 1.0939 11.520 1.0000 5.6000 1.0933 11.680 1.0000 5.7600 1.1025 11.840 1.0000 5.9200 1.1149 12.000 1.0000 l

i Page 17 of 27

(

1

4 EXAMPLE

(

~

1.5 1

1.45 1.4 1

1.35 1.3 s

a b

f 1.25 c

a 8

z, x

s 6

1.2 e

a A

a g)W 1.15

.d ac A'

r._.

1.1 1.05 l

1 0

1 2

3 4

5 6

7 8

9 10 11 12 CORE HEIGHT (FEET) 1 FIGURE 10 V. C. SUMMER RAOC W(Z) AT 16000 MWD /MTU l

Page 18 of 27 i

u_ __ __ _

1

~ '

EXAMPLE

\\

DATA FOR FIGURE 10 V. C. SUMMER RAOC W(z) AT 16000 MWD /MTU l

Core Heicht W(z)

Core Heicht W(z) i

)

0.0000 1.0000 6.0800 1.2179 i

0.1600 1.0000 6.2400 1.2281 0.3200 1.0000 6.4000 1.2371 0.4800 1.0000 6.5600 1.2445 0.6400 1.0000 6.7200 1.2504 0.8000 1.0000 6.8800 1.2548 0.9600 1.0000 7.0400 1.2577 t

1.1200 1.0000 7.2000 1.2590 1.2800 1.0000 7.3600 1.2587 1.4400 1.0000 7.5200 1.2568 1.6000 1.0000 7.6800 1.2535 1.76 @

1.2672 7.8400 1.2486 1'sth 1.2484 8.0000 1.2425 2.082J 1.2291 S.1600 1.2347 2.2400 1.2094 8.3200 1.2261 2.4000 1.1898 8.4800 1.2221 2.5600 1.1702 8.6400 1.2222 2.7200 1.1487 8.8000 1.2201 2.8800 1.1301 8.9600 1.2173 3.0400 1.1234 9.1200 1.2170 3.2000 1.1240 5.2800 1.2227 3.3600 1.1265 9.4400 1.2342 3.5200 1.1283 9.6000 1.2503 3.6800 1.1320 9.7600 1.2683 3.8400 1.1393 9.9200 1.2866 4.0000 1.1461 10.080 1.3047 4.1600 1.1518 10.240 1.3236 4.3200 1.1569 10.400 1.0000 4.4800 1.1613 10.560 1.0000 4.6400 1.1647 10 720 1.0000 4.8000 1.1670 10.880 1.0000

4. % 00 1.1686 11.040 1.0000 5.1200 1.1689 11.200 1.0000 5.2800 1.1677 11.360 1.0000 5.t400 1.1700 11.520 1.0000 5.6000 1.1792 11.680 1.0000 5.7600 1.1924 11.840 1.0000 5.9200 1.M61 12.000 1.0000 Page 19 of 27

EXAMPLE 1.15 1.14 1.13 1.12 1.11 N%sq 1.09

^4, 1.08

\\

S

'a 3

6 E

1,07 6

.a

't.

f bg

'a 1.06 f

a' a

6'j 1.05 "A

e" 1.04 1.03 1.02 1.01 1L 0

1 2

3 4

5 6

7 8

9 10 11 12 CORE HEIGHT (FEET)

FIGURE 11 V. C. $UMMER BASELOAD W(Z) AT 150 MWD /MTU Page 20 of 27

____..____-__o_,

_W

EXAMPLE DATA FOR FIGURE 11 V. C. SUMMER BASELOAD W(z) AT 150 MWD /MTU Core Heiaht Wiz)

Core Heicht W(z) 0.0000 1.0000 6.0800 1.0582 0.1600 1.0000 6.2400 1.0554 0.3200 1.0000 -

6.4000 1.0528 0.4000 1.0000 6.5600 1.0511 0.6400 1.0000 6.7200 1.0495 0.8000 1.0000 6.8800 1.0477 0.9600 1.0000 7.0400 1.0465 1.1200 1.0000 7.2000 1.0438 l

1.2800 1.0000 7.3600 1.0432 1.4400 1.0000 7.5200 1.0445

} '~

1.6000 1.0000 7.6800 1.0466 1.7600 1.0987 7.8400 1.0483 1.9200 1.0985 8.0000 1.0500 2.0800 1.0982 8.1600 1.0517 2.2400 1.0979 8.3200 1.0533 2.4000 1.0974 8.4800 1.0550 2.5600 1.0968 8.6400 1.0568 2.7200 1.0961 8.8000 1.0588 2.8800 1.0952 8.9600 1.0609 3.0400 1.0941 9.1200 1.0630 3.2000 1.0929 9.2800 1.0650 3.3600 1.0915 9.4400 1.0669 3.5200 1.0900 9.6000 1.0688 t

3.6800 1.0884 9.7600 1.0706 f

3.8400 1.0868 9.9200 1.0723 4.0000 1.0852 10.080 1.0739 4.1600 1.0836 10.240 1.0754 4.3200 1.0819 10.400 1.0000 4.4800 1.0801 10.560 1.0000 4.6400 1.0782 10.720 1.0000 4.8000 1.0762 10.880 1.0000 4.9600 1.0741 11.040 1.0000 5.1200 1.0718 11.200 1.0000 5.2800 1.0694 11.360 1.0000 5 4400 1.0672 11.520 1.0000 5.6000 1.0651 11.680 1.0000 5.7600 1.0629 11.840 1.0000 5.9200 1.0606 12.000 1.0000 Page 21 of 27

EXAMPLE 1.15 1.14 1.13 1.12

'd 1.11 1.1 h

a a

1.09 a

a

^

1.08 a

c a

k

'a 1.07 pmm

/

1.06 3.

"t.

d' p

'de 1.05 1.04 1.03 1.02 1.01 1

0 1

2 3

4 5

6 7

8 9

10 11 12 CORE HEIGHT (FEET)

FIGURE 12 V. C. SUMMER BASELOAD W(Z) AT 8000 MWD /MTU l

l l

Page 22 of 27 l

l l

t EXAMPLE DATA FOR FIGURE 12 V. C. SUMMER B ASELOAD W(z) AT 8000 MWD /MTU j

Core Heicht W(z)

Core Heicht W(z) 0.0000 1.0000 6.0800 1.0555 0.1600 1.0000 6.2400 1.0539 0.3200 1.0000 6.4000 1.0534 0.4800 1.0000 6.5600 1.0538-0.6400 1.0000 6.7200 1.0545 0.8000 1.0000 6.8800 1.0556 0.9600 1.0000 7.0400 1.0575 1.1200 1.0000 7.2000 1.0595 1.2800 1.0000 7.3600 1.0612 1.4400 1.0000 7.5200 1.0625 1.6000 1.0000 7.6800 1.0637 1.7600 1.1139 7.8400 1.0648 1.9200 1.1121 8.0000 1.0658 2.0800 1.1100 8.1600 1.0666 2.2400 1.1075 8.3200 1.0674 2.4000 1.1048 8.4600 1.0680 2.5600 1.1018 8.6400 1.0685 2.7200 1.0985 8.8000 1.0687 2.8800 1.0951 8.9600 1.0687 3.0400 1.0914 9.1200 1.0687 3.2000 1.0873 9.2800 1.0689 3.3600 1.0833 9.4400 1.0699 3.5200 1.0806 9.6000 1.0719 3.6800 1.0787 9.7600 1.0739 3.8400 1.0769 9.9200 1.0755 4.0000 1.0749 10.080 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 4.8000 1.0683 10.880 1.0000 4.9600 1.0C74 11.040 1.0000 5.1200 1.0663 11.200 1.0000 5.2800 1.0651 11.360 1.0000 5.4400 1.0636 11.520 1.0000 1

5.6000

' 0618 11.680 1.0000 5.7600 1.0600 11.840 1.0000 l

5.9200 1.0579 12.000 1.0000 l

l Page 23 or 27

~

EXAMPLE 1.15 3

2.

1.14 a

^

1.13 a

1.12 f

3 a

^

1.11 o

s t.

1.1 6

.s a

1.09 o

a 1'08 E

'A

/

I

-4' 1.07

^

6 6

1.06

^

a f u

1.05 1.04 1.03 1.02 1.01 1

0 1

2 3

4 5

6 7

8 9

10 11 12 CORE HEIGHT (FEET) l FIGURE 13 V. C. SUMM ER B ASELOAD W(2) AT 16000 MWD /MTU 1

Page 24 of 27

EXAMPLE DATA FOR FIGURE 13 V. C. SUMMER BASELOAD W(z) AT 16000 MWD /MTU Core Heicht W(z)

Core Heicht W(r) 0.0000 1.0000 6.0800 1.0616 0.1600 1.0000 6.2400 1.0647 0.3200 1.0000 6.4000 1.0678 0.4800 1.0000 6.5600 1.0706 0.6400 1.0000 6.7200 1.0730 0.8000 1.0000 6.8800 1.0752 0.9600 1.0000 7.0400 1.0771 1.1200 1.0000 7.2000 1.0787 1.2800 1.0000 7.3600 1.0800 1.4400 1.0000 7.5200 1.081'1 1.6000 1.0000 7.6800 1.0820 1.7600 1.1488 7.8400 1.0826 1.9200 1.1436 8.0000 1.0830 2.0800 1.1380 8.1600 1.0832 2.2400 1.1321 8.3200 1.0833 2.4000 1.1259 8.4800 1.0830 2.5600 1.1193 8.6400 1.0838 2.7200 1.1123 8.8000 1.0880 2.8800 1.1050

8. % 00 1.0932 3.0400 1.0976 9.1200 1.0976 3.2000 1.0897 9.2800 1.1019 3.3600 1.0828 9.4400 1.1060 3.5200 1.0799 9.6000 1.1098 3.6800 1.0787 9.7600 1.1133 3.8400 1.0766 9.9200 1.1165 4.0000 1.0744 10.080 1.1192 4.1600 1.0722 10.240 1.1217 4.3200 1.0699 10.400 1.0000 4.4800 1.0675 10.560 1.0000 4.6400 1.0651 10.720 1.0000 4.8000 1.0626 10.880 1.0000 4.9600 1.0600 11.040 1.0000 5.1200 1.0574 11.200 1.0000 5.2800 1.0557 11.360 1.0000 5.4400 1.0557 11.520 1.0000 5.6000 1.0571 11.680 1.0000 5.7600 1.0580 11.840 1.0000 5.9200 1.0591 12.000 1.0000 Page 25 of 27

l EXA MPLE

~Cora Oparating Limits Report

v. c. s-c-2.6 RCS Flow Rate and Nuclear Enthalov Rise Hot Channel Factor - FN (Specification 3.2 3)

FN R=

RTP e

FaH (j, pF H * (1-P))

A THERMAL POWER where P=

RATED THERMAL POWER 2.6.1 FaHRTP= 1.56 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 Figure 14.

1 Page 26 of 27

EXAMPLE MEASUREMENT UNCERTAINTIES OF 2.1% FOR FLOW AND 4.0% FOR INCORE MEASUREMENTOF FN AHARE QNCLUDED IN THIS FIGURE 38 I

I ACCEPTABLE UNACCEPTABLE OPE RATION RE GION OPERATION REGION 36 34

'W E

-0 32 h

t Od

(

30 5

&y

(("[,

(1.00.28.95)

cc (1.00,28 66)

(1.00.28 37) 28 (i 00.28 Os)

SEE NOTE (1 00.28 79)

(1.00,27 50,

26 24

.9

.95 1

1.05 1.1 R=FNan/1.56[1.0 + 0.3(1.0.P)]

NOTE: When operating in this region, the restricted power levels shall be considered to be 100% of rated thertaal power (RTP) for Technical 5pedrfication Figure 2.11 FIGURE 14 RCS TOTAL FLOW RATE VERSUS R FOR THREE LOOP OPERATION Page 27 of 27 k

\\

_--__o

,o No Significant Hazards Evaluation

(-

to Document Control Desk' Letter

,+

September 19, 1989 Page 1 of 4 NO SIGNIFICANT HAZARDS EVALUATION FOR CHANGES TO TECHNICAL SPECIFICATIONS DELETING CERTAIN CYCLE-SPECIFIC PARAMETERS Backcround Generic Letter 88-16, 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 Operating Limits Report (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 burden 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, future Virgil C. Summer Nuclear 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 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.

Proposed Chance The proposed technical specification changes concern the relocation of several cycle-specific core operating limits for Virgil C. Summer Nuclear l

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 !which 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.

' ' to Document Control Desk Letter I, !.

September 19, 1989 o

Page 2 of 4 L

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 (c).

3.1.3.6 Control Rod Insertion Limits (d) 3.2.1 Axial Flux Difference (e).

3.2.2 Heat Flux Hot Channel Factor

.(f) 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 proposea 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 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).

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 values determined to be within the acceptance criteria using an NRC-approved

. calculation methodology. As previously discussed, the methodologies for 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 has no impact upon plant operation or safety. No safety-related equipment, safety function, or plant operations will be altered as a result of this l

proposed change. Since the applicable FSAR limits will be maintained and the Technical Specifications will continue to reouire 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.

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 Carolina Electric & Gas Company calculations, including mid-cycle or other

l

- Attachment 4 to Document Control Desk Letter b&

September 19,'1989 Page 3 of 4 I

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

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 appropriate core operating limits for the applicable cycle will apply.

Technical Specifications will not be changed.

Determination of Significant Hazards Pursuant to 10 CFR 50.91, South Carolina Electric & Gas Company has determined that operation of the facility in accordance with the proposed license amendment request does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR 50.92. The following discussion describes how the proposed amendment satisfies each of the three standards of 10 CFR 50.92(c).

1)

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The removal of cycle-specific core operating limits from the Virgil C.

Summer Nuclear Station Technical Specifications has no influence or impact on the probability or consequences of any accident previously evaluated. The cycle-specific core operation limits, although not in Technical Specifications, will be followed in the operation of the Virgil C. Summer Nuclear Station. The proposed amendment 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 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 of 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 evaluated.

2)

The proposed change does not create the possibility of a new or 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 his proposed change. The cycle specific variables are calculated (usingtheNRC-approvedmethodsandsubmittedtotheNRCto allow the Staff to continue to trend the values of these limits. The Technical Specifications will continue to require operation within the required core operating limits and appropriate actions will be taken when or if limits are exceeded.

6 c I C Attachment'4 to Document Control. Desk Letter

.1 September'19, 1989

.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-core 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.

Conclusion

The Commission has provided guidance concerning the application of the standards for~ determining whether a significant hazards consideration exists.

This guidance (51 FR 7750) includes examples of the type of amendments 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 determined that the proposed license amendment does not involve any significant hazards considerations.

N

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ML20247L728

Text

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