Text

Proposed Tech Specs,Including Page 1-6c,re Relocating cycle-specific Core Operating Limits
	ML20245C235
Person / Time
Site:	Davis Besse
Issue date:	06/16/1989
From:	; TOLEDO EDISON CO.
To:	;
Shared Package
ML20245C209	List: ML20245C208; ML20245C230; ML20245C235;
References
	NUDOCS 8906260101
	Download: ML20245C235 (42)
	v • d • e

- - - - - - - _

' ' Docir$tNumbar50-346 License Dumber NPF-3

' Serial Number 1668 Page 4 AfDEX ll DEFINIff0NS SECTION PAGE 1.0 DEFINITI0MS (Continued)

SOLI DI FI CATI ON........................................

1-6a OFFSITE DOSE CALCULATION MANUAL (0DCM)................

1-6a GASEOUS RADWASTE TREATMENT S YSTEM.....................

1-6a VENTI LATI ON EXHAUST. TREATMENT S YSTEM..................

1-Ga 3

PURGE-TURGING.........................................

1-66 VENTING...............................................

1-6b MEMB ER (S ) 0 F THE PUB LI C...............................

1-6b S_ITEB00HDARY.........................................

1-6b UNRESTRICTED AREA.....................................

1-6b D EWAT E R I N G............................................

l.-6b r

OPERATIONAL MODES (TABLE 1.1).........................

1-7 FREQUENCY NOTATION (TABLE 1. 2)........................

1-6

~m CaecmAm bam E" "

%c vd) vv. t i

l 1

1 DAVIS-BESSE. UNIT 1 Ia Amendment No. E5 1

h f

J I

P

l

. Dock,et Number'501346 Lit:n;2 Number NPF-3 L

52 rial Number 1668

' Attachment 1 P ge:5 DEFINITIONS e we.- -

_y _

CORE OPE RATtMG b(MtTslEPoRT l,4 l %c Core OpeRArix: l.iuiTrTemer is EL, ami1-sp,c,-(c documeal. Ad. provides cove operating mits f r the curreal li p

reload cycle.%ese cycle spcifik core gerab li y

mits shaR be determiaca for eact, reload eqcle in accordace' win f 3 ectpcation 6.9.1.1.7tamt c>per % wimm Rese core p

op,eatig imits is. addressed in individual s7ecifcabs.

l v--

i DAVIS-BESSE, UNIT 1

' Dock'et Nuxber 50-346

, c p-

& Liccnra Nurber NPF-3 S rial Number 1668 P:ge-6 REACTIVITY C0!: TROL SYSTEw.S MODERATOR TEMPERATURE COETTICIEh7 LIMITING CONDITION FOR OPERATION m, -.

3.1.1.3 dhg madsrutter e-7 -entrea p - M4r4-ne_ N gi e b1ainhNed w h dw w. etape G its prov.aca Us e CoRECPERATtW, L18iTs EPORT'.

3.'fi.ess posa lve tnan 0.y = 10-" tk/k/*F whenever THER.4AL P0kTR is < 95%1f'

(

RAAr b m.tQd.AL POWER,

/

(

i

"^s f.4 Less positive than~0.0'?-1 F 7 henever THERMAL POWER is 2 95% of

} RATED THERMAL P0kTRdiuf ht

[

t c-regia 1 to or less negative than -3.0 x 10~".tk/k/* nam TEEFEAL _POWERp APPLICABILITY: MODES 1 and 2*#.

ACTION:

With the moderator temperature coefficient outside the above limits, be in at least HOT STANDBY vithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMEh75 4.1.1.3.1 The MTC shall be detettined to be withiniftsilli3ts by confirmatory l

measurements. MTC measured values shall be extrapolated and/or compensated to per=1t direct comparison with the above li=its.

4.1.1.3.2 The MTC shall be determined at the following frequencies and I

THERV.AL P0kTR conditions during each fuel cycle:

a.

Prior to initial operation above 5* of RATED THERMAL P0kTR, af ter each fuel loading.

1 b.

At any TEEEMAL PokTR, within 7 days af ter reaching a RATED THEL".AL P0kTR equilibri'.:= boren concentration of 300 ppm.

I.

l I

With k,ff 2 1.0.

See Special Test Exception 3.10.2.

DAVIS-BESSE,'JIT 1 3/4 1-4 Amendment No. 45

n c

1 DocNe'tNumber50-346-

'Lic:n n Nh;ber NPF-3

'S: rial Number 1668 ADDlil0NAL CHANGES PREVIOUSLY

", f *"

PROPOSED By LETTER p

Serial No. I@7 Date H/2 /87 REACTIVITY CONTROL SYSTEMS ACTION:

(Continued) i c)

A power ' distribution mgp is obtained from the-incore i{

d detectors and F and F are their 1imits wiShin 72 $ours. verified to be within 0

d)

Either the THERMAL POWER level is reduced to < 60%

of the THERMAL POWER allowable for the reactor _

coolant pump combination within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Flux Trip Setpoint is i

reduced to < 70% of the THERMAL POWER allowable for the reactor coolant pump combination, or e)

The remainder of the rods in the group with the inoperable rod are aligned to within + 6.5% of the k,5 YY*h9 M E I5 T 5 Or. U _. N t5e'YHYR kl.-NUiNk i

s.na e res r1cted pursuant to Specification 3.1.3.6 during subsequent operation.

&( b redy witW acceptable opudi g posifiono limils SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be determined to be within the group average height limit by verifying the individual rod positions 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 Asyninetric Rod Fault Circuitry is inoperable, then verify the indi-vidual rod position (s) of the rod (s), with inoperable faul Circuitry at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1. 3.1. 2 Each control rod not fully inserted shall be determined to i

i be OPERABLE by movement of at least 2% in any one direction at least once every 31 days.

l DAVIS-BESSE, UNIT 1 3/4 1-20 i

'l

' Dockst Nu:ber 50-346 Lice.nsa Numb:r NPF-3 S rial Number 1668 Paga 8 REACTIVITT C0!TTROL SYSTEMS REGULATING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3M_hatingJojlgg0pps shall beflfeited in ph,.ic.1 rn i

5.

= Figures-3rl-C.ud -2L, 36 Se,.nd -3b.

f.-ro]d i.: n t i ait-a :

greep e' eri p of 2515% sh='l be sinteined L6.--a.equential.-ithd avn )j g en-- 5, < - 2 ps; Ein e a wi thin h sucp.ble r-o e, a bi,,

O~i ts yw J rea u tat in, rod desikk. emvid ed 6 d e Co4E CPERATitJG L yr,s R op7 APPLIYABY1NT N 0f$5 ACTION E

Vith the regulating rod groups inserted beyond thesheAr4>

l limits (in a region other than agcce table operation), or with any group sequence or overlap outside the g(Spgiticar4nn LI A12 g iTi F,/ limits, except for l

surveillance testing pursuant to s

, either auotable coe. ti,9f Restore the regulating groups M tile-dimits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , a.

or 3

l b.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THER!iAL POVER vhich is allowed by the rod group positioruMhe

< t$c /:"fWvithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or lia." ts u

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

c.

NOTE:

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

See Special Tes t Exception 3.10.1 and 3.10.2.

tVith k,gf 1 1.0.

DAVIS-BESSE. UNIT 1 3/4 1-26 Amendment No. 11 33, 41, 42, 45, 61, 69, 80,123

7-

r; ci

.'Dockht1Nuber)50-346-

}Lic:nsinumberNPF-3 i

Serial-Number 1668

['

lAttachment 1

.l Page 9-1 l

)

REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS

?

SURVEILLANCE REQUIREMENTS' Ac.ceph55 i

iophh requiatin group shall'be determined to be 4.1. 3. 5. T os within th

--- Len,_1=Rn_ce add _ ccr,a; limits at least' once every hen:

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months excep w

a..

The regulating ro'd insertion limit alann is' inoperable, then.

verify the groups to be within the. insertion limits'at least once.per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; b.,

The control rod drive sequence alarm.is inoperable,'then verify.the groups to be within the sequence and overlap limits at least' once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

DAVIS-BESSE, UNIT 1 3/4 1. 27 a

Docket Nurber 50-346 l

'ILicen3'Nu ber NPF-3 S; rial Number 1668 3

? Attachment'l-Pcge 10 Figure 3.1-2a Regulating, Group Position Limits,

/

\\

0 to 325 10 EFPD, Four RC Pumps --

Davis-Besse 1, Cycle 6

./

D

'N.

\\

(300,102)

(258,102) 100 - P6wer Level 3

Cut'off = 100%

(270,102)

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(270,92) i

'N SHUTDOWN E

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MARGIN g 80

.s LIMIT (250,80) c-N J

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w' UNACCEPTABLE N

'A OPERATION

\\,s OPERATION 60

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RESTRICTED

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C 70,50%'

(180.50)

.o ea 5 40 j/-

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l g

g

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(128,283)

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b

/

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$ 20 s

. ACCEPTABLE

/

'0PERATION 4

(0.0,5.0)/

\\

0 0

/

100 200

\\

300 Rod Index (% Withdrawn)

\\

GR 5 i i

i 0

75 iu:)

N, 1

GR 6 i

i i

1 0

25 75 100 GR 7 i 0

25 10 1

\\

i N

\\

-i DAVIS-BESSE, UNIT 1 3/4 1-28 Amendment No. 11, 33, 45, '

61, 80, 123

EE i.

1 x.,

';.... Dock 3t Number 50-346

~

)'Lic';nsa Nu;ber NPF-3 i

- 5: rial Number 1668

)

P ge 11 N..

Figure 3.1-2b Regulating Group Postion Limits After 325 10 EFPD, Four RC pumps, APSRs Withdrawa - Davis-Besse 1, Cycle 6 j

'y wM

\\

(266,102 (300,102)-

\\

100

'q Power Level

, (270,102) xCutoff = 100%

( 0,92)

SHUTDOWN

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MARGIN (250,80)

\\

LIMIT 80

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E i

s OPERATION

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RESTRICTED

=

60 UNACCEPTABLE s

OPERATION 1

(17.50)

(180,50) e j

40 p

E f

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(136';28.5)

E

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f 20

/

ACCEPTABLE

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OPERATION i

(0.0,5.0,

's 0

J 0

/

100 200

'N 300

/

Rod Index (*.' Withdrawn)

N

/

\\

GR 5 '

i 0/

75 100

\\

/

GR 68 0

25 75 100 GR 7 y

0 25 x100 f

s

/

'N

{

,/

x

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's 1

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/

DAVIS-BESSE UNIT 1 3/4 1-28a Amendment No. 11, 33, 42 \\

45, 61, 80, 123 4

r

'c Do'ket Number 50-346 bicench Number NPF-3 Serial Number 1668

/

Page 12 jr

}

\\

Figure 3.1-3a Regulating Group Position Limits,

/

\\

0 to 325 10 EFPD, Three RC Pumps --

Davis-Besse 1, Cycle 6

.x

/'

/

1 100 y

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\\

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258,77) 00,77) 80 2

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SHUTDOWN

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MARGIN (270,69.5)

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LIMIT l

5 60

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(250,60.5)

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y 40 8

(180,38) l b

a.

i 50

/

2 4

(128,21.8)

ACCEPTABLE

\\

OPERATION s

0 0

/

100 200 300 Rod Index (% Withdrawn)

/

GR 5 '

0

/

75 100 GR 6 i

\\

/

0 25 75 100 s

GR 7 '

0 25 100

/

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l

/

DAVIS-BES5?. UNIT 1 j'

3/4 1-29 Amendment No. 11, 33, 41,\\\\

45, 61, 80, 123

/

)

i--:b' ':

Docket Nurber 50-346

, Lic:nsa Nuzber NPF-3

' ^

Serial. Number 1668.

i

Attachment 1 Page 13 N

Figure 3.1-30 Regulating Group Position Limits N

After 325 10 ETPD, Three RC Pumps, APSRs Withdrawn -- Davis-Besse 1, r

Cyc d p

N.

3ELETGb

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- 100 -i

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(300,77) s M

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MARGIN

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g UNACCEPTABLE LIMIT

/

cc OPERATION (270,69.5)

E 60 N

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S Q

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OPERATION y'(j

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RESTRICTED 40 a

5 (180,38)

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20 (136,21.8) e ACCEPTABl.E s

/

GPERATION

'(0,4.25 )

/

0

/

100 200 N

300

)

/

Rod Index (% Withdrawn) i GR 5 0

75 100 GR 6 '

0 to to 100 GR 7 '

0 25 100

/

g DAVIS-BESSE, UNIT 1 3/4 1-29a Amendment No. 11, 33, J.2, \\

45, 61, 80,123

\\

j

' ~' '

bockat Nurbar 50-346

.L c nsa Number NPF-3 i

+ :

Serial Number 1668 4

Page 14?

L REACTIVITY CONTROL' SYSTEMS l.

R00 PROGRAM-LIMITING CONDITION FOR OPERATION e-_=

NQ

'M

?,

3.1.3.7 Each control-re safety S M and APSR)'shall be pro--

gammedt o erate in t ore _.

_ land rod group specified in CeRECPERATme. Lmirs TEPoRT. '

'%.iiere s.,.-4.

w J

APPLICABILITY: MODES 1* and 2*.

ACTION:

With any control rod not programmed to operate as specified above, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLANCE REQUIREMENTS 4.1.'3.7 a.

Each control rod shall.be demonstrated to be programed to operate in the specified core position and rod group by:

1.

Selection and-actuation from the control room and verifi-cation of rmvement of the proper rod as indicated by both the absolute and relative position indicators:

a)

For.all control rods, after the control rod drive patches are locked -subsequent to test, reprogramming or maintenance within the panels.

b)

For specifically affected individual rods, following maintenance, test, reconnection or modification of power or instrumentation cables from the control rod drive control system to the control rod drive.

2.

Verifying that each cable that has been disconnected-has been properly. matched and reconnected to the specified 3

control rod drive.

j b.

At least once each 7 days, verify that the control rod drive patch panels are locked.

See special Iest Exceptions 3.10.1 and 3.10.2.

DAVIS-BESSE, UNIT 1 3/4 1-30 Amendment No.11

)

1

' Dock $tNurbar50-346 Lictn:2 Nurber NPF-3 S: rial Number 1668 Figure 3.1-4 Cait.o1 Rod Core Locations

' Attachment 1 and Group Assignments --

-Pege-15 Davj s-Besse 1, Cycle 6 y

A

.N A

\\

g

'N 4

6 4

C

\\

2 5

5 l2

/

D

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7 8

7

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\\

2 S

5 f

E F

4

\\

8 6

3 6

8/

4 G

S \\.

1 1

/

i, H

W-

,6 7-3 4

3 /

7 6

-Y x

K 5

1 1 /

5 L

4 8

6x 3

/

6 8

4 M

2-5

\\

/

5 2

N l

7 8

(

8 7

o l

2 3

5 2

s P

l l

l

,F 6

\\

4 R

i i/

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N\\

Z k

13 14 15 1

2 3

4 5

6 7

8 9

10 11 Fun \\

Group No. of Rods ction 1

4 Safet'ys 2

8 Safety.

3 4

Safety \\

X Group Number 4

9 Safety 5

12 Control 6

8 Control 7

8 Control 8

8 APSRs N

\\

Total 61 s

DAVIS-BESSE. UNIT 1 3/4 1-31 Amendment r:o. 11, 33, 45, \\

61, 80, 123

\\

5 4

JDsckst.Nunber 50-346i

License Number NPF-3

> Serial Number 1668~

' Attachment 1 3

' Page 16 REACTIVITY CONTROL SYSTEMS XENON REACTIVITY LIMITING CONDITION FOR OPERATION THER%L 29WER shali.not be increased 'above the power level cutoff -

3.1.3.8 specified'i ?.g,ur: f C Q h u f +he ellowing y l

satisfied:.

th'**uetAue am~u3 IW as Yo - re%ueiy ro psiha p' ici.J l

in %, CORE C)?Efi ATsuG L a H irs R EPoRT*, J _

^

l Xenon. reactivity is within 103Mthe equilibrium a.

value for RATED THERMAL POWER and is approaching stability, or b.

THERMAL POWER has been within a range of 87 to 92 percent -

of RATED l THERMAL POWER for a period exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in the-soluble poison control mode, excluding xenon free start-ups.

l I

APPLICABILITY: MODE 1.

ACTION:-

With the requirements of the above specification not satisfied, reduce THERMAL' POWER to less than or equal to the power level cutoff within 15 minutes.

SURVEILLANCE REQUIREMENTS Xenon reactivity shall be detennined to be within 10% of.the 4.1. 3. 8 equilibrium value for RATED THEPMAL POWER and to be approaching stability or it shall be determined that the THERMAL POWER has been in the range of 87 to 92% of RATED THERMAL POWER for > 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , prior to increasing THERMAL POWER above the power level ciitoff.

DAVIS-BESSE, UNIT 1 3/4 1-33

Dock,et Nusber 50-346

'Lic nsa Nu-ber NPF-3

5:4i01 Nu ber 1668

.)

I Page 17 REACTIVITT CONTROL' SYSTEMS AXIAL POVER SHAPING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION

-w 3.1.3.9 The axial power shaping rod group shall Mmi!M in phy ieel E'-h n. Piswes 11-56 -5t, an. -.c.

Win % < h' Pra bic opeah g its for **

r pe d Et,2 N U., Cc R E CPE R Af g hic llM i rs RE Pe.m i.

APPO M ITY,:

MODES 1 and 2*.

~ ' - ~

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

Restore the axial power shaping rod group to within the limits a.

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

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

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

l

\\

1 SURVEILLANCE REQUIREMENTS The position of the%[jEp'tible o x1Mr shaping rod group shall be 4.1.3.9 determined to be within thel ola limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except when the axial,/jg?a' ping rod insertion limit alarm is l

pover s inoperable, then verify the group to be vithin th 1:i:~rilN g l

at least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Acce p t Al e o pe ra ti.,.

,"Wowbui)

Vith K,gf 1 1.0.

DAVIS-BESSE. LUIT 1 3/4 1-34 Amendment No. 33.

42, 45, 61, 69.

80,123

'-D ekst Nu ber.,50-346 1*ic:ns5NurbarNPF-3 Serial Number 1668 Attachment ~1 Page 18 Figure 3.1-Sa APSR Position Limits,.0 to 3251.10 EFPD, Four FCJumpg-Davis-Besse 1, Cycle 6 s

p

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f 3ELETE S

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RESTRICTED REGION

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100 P 0,102)

(100,102) 'z

s..

N

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N 80

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s N

N

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a b

60 "j

P RMISSIBLE OPERATING REGION l

ad I

..o 40

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N g

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20

'4 j

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10 20 30 40 50 50 70 30 90 100 j

s

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APSR Positi0n (?. Withcrawn)

Ns T\\

l N

i DAVIS-BESSE, UNIT 1 3/4 1-35 Amendment No. 33, 45, 61,

/

30,123 d

.r I'-

'- ' " - !Lic;n a Nu ber NPF-3

'Do'ckst Nu;ber 50-346 f

L. Serial Number 1660

.-Attachment 1

'Page 19

\\

Figure 3.1-Sb APSR Position Limits After

.\\.

3251 10 EFPD, Three 'or four RC

,e

\\

Pumps, APSRs Withdrawn --

\\

Davis-Besse'1_, Cycle 6

\\

100

'N

/

E - 80 x

5

N 2

N APSR INSERTIONA 0T ALLOWED W

60 A-IN THIS T INTERVAL S

E

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40

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N

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20

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2

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'90 100

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APSR Position (". Withdrawn)

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s y

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l N

l s

s

/

i DAVIS-BESSE, UNIT 1 3/4 1-36 Amendment No. 33, 42, 45 s

\\

/

61, 80.123

7, l

k',:DocloetNurber50-346

' 1. Lic:n;p Nu;Mr NPF-3

)

S: rial Number 1668

Attachment l'

'P2ge 20

\\.

. Figure 3.1-5c.. APSR Position Limits, O to 325t 10 EFPD,.

\\.

Three RC Pumps -- Davis-Besse 1, Cycle 6

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J.MLE i Le-7

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RESTRICTED REGION 2

W (0,77)

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(100,77) 1

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60 S

w

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5

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PERMIS$1BLE 40 5

OPERATING REGION E

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E

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N u

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20

/

N

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9 i

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10 20 30 40 50 60 70 30

)30

00 APSR Position (* Withdrawn) 1 i

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J N

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l DAVIS-BESSE, UNIT 1 3/4 1-37 Amendment No. 33, 45, 61

/

j

69. 80,123

'N j

I

\\.. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _.

Dociat Nuxbtr 50-346 Licenza Nuzbar NPF-3 S: rial Number 1668 Page 21 3/4.2 POVER DISTRIBUTION LIMITS AXIAL POVER IMBALANCE LIMITING CONDITION FOR OPERATION 3 a Q XIAL J )RE JRBALANEshall_be maintained _Vithin_theJ,4mi4s-shown 1 1

fon F4 guns 3.2-1 and 3.2 2. acc Pt*

Axi A' ec WG R In BAL AVc k opentiny Lits /

rvv.ded in the ORE CPERATsWq LlhiTS 'RE FORT.

g-APPLI'C311 TIIM'0YiE'l above 40%

TED THERMAL POVER.*

ACTION Vith AXIAL POVEE IMBALANCE exceeding the limits specified above, either:

Restore the AXIAL POWER IMBALANCE to withi,nd Q {mits_vithin 15 a.

minutes, or a cc'<>ta uc o p mem.

v b.

Vithin one hour reduce power until# mbalance limits are met or to 40% of RATED THERMAL POWER or lessi g,,

c SURVEILLANCE REQUIREMENTS g y-.-

La ccep ta tJ< ownw.

4.2.1.

The AIIAL POVER IMBALANCE shall be determined to be withirtv,

limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when above 40% of RATED THERMAL" POVER except when the AXIAL POVER IMBALANCE alarm is inoperable, then calculate the AXIAL POWER IMBALANCE at least once per hour.

See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-1 Amendment No. 33, 42, 45, 61, 69, 80,123

!Dockst.Nd:bbr 50-346

..-.,Lic;nco Numb:r NPF-3 Serial Number 1668 l

- Page 22 Figure 3.2-1 AXIAL POWER IMBALANCE Limits,

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61, 80, 123 l

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".LDoc'k$t Nu;ber 50-346 Liesnca Nuzber NPF-3

' Y Serial Number 1668'

' Attachment 1 Page 23

'N Figure 3.2-2 AXIAL POWER IMBALANCE Limits,

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DAVIS-BESSE, UNIT 1 3/4 2-3 Amendment No. 11, 33, 45, 61, 80, 123

Docket Nu;ber 50-346-

' Licdc2 Number NPF-3 L

S; rial Nu:ber 1668

[

.Attcchment 1 P:g3 24 1

POVER DISTRIBUTION LIMITS QUADPANT POVER TILT LIMITING CONDITION FOR OPERATION

/32d l!DLMADEANL20m TTLTahalLngl_. exceed 4t_Slaa_dy_ State _ ueq 4 2-TsL16 3.2,-1. k e QCh Miw T t bGO' TLT f(c M/ed nr % a/M oby/>riq

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QtM1/ rA d%fcil" W APPMI'TM,DE 1 above 15% of RATED THERMAL POWER

i i

ACTION:

Vith the QUADRANT POVER TILT determined to exceed the Stea.dy.~

a.

Sta,Ie_Limij b,Lut. Jess _than or equal 3Je Transient _Limi_thf- $

debrie 4.2-1. m tu ce u: cf62Mr W6 '/d/h 57127/

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1.

Vithin 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

l a)

Either reduce the QUADRANT POWER TILT to within its Steady State Limit, or b)

Reduce THERMAL POVER so as not to exceed TUERMAL POVER, including power level cutoff, allovat.le for the reactor coolant pump combination less at least 2%

for each 1% of QUADRANT POVER TILT in excess of the Steady State Limit and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , reduce the High Flux Trip Setpoint and the Flux-6 Flux-Flow Trip Setpoint at least 2% for each 1% of OUADRANT POVER TILT in excess of the Steady State Limit.

2.

Verifv that the QUADRANT POVER TILT is within its Steady Stat ~ Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the Steady State Limit or reduce THERMAL POVER to less than 60% of THERMAL POVER allowable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to $ 65.5% of THERHAL POVER allovable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POVER OPERATION above 60% of THERMAL POVER allovable for the reactor coolant pump combination may proceed provided that the QUADRANT POVER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until erified acceptable at 95% or greater RATED THERMAL POVER.

See Special Test Exception 3.10.1 DAVIS-BESSE. UNIT 1 3/4 2-9 Amendment No.123 l

l

I Dot)tetlNunher 50-346 4

J

"' ;Licensa, Number NPF-3 Serial Number 1668 A00iitoIML CHANGES PREVIOUSLV' s

! Attachment 1

-. PROPOSED SY LETTER Page 25' SerialNo. I'/01 '

Date///2./av,

POWER DISTRIBUTION LIMITS

' LIMITING CONDITION FOR OPERATION (Continued)'

b..

Vith the QUADRANT POVi.'R TILT determined ' to exe he Transient

~

Limit but less than. the Maximum Limit f ~1:h 2.L*, due to l

misalignment.of either a safety, regulatin [~r'axaipower-.

shaping rod:

G72 cWMoN4 4

avWs daMr 1.

Reduce THERMAL POVER at least 2% for each 1% of indicate QUADRANT POVER TILT in excess of the Steady State Limit

'vithin 30 minutes.

-t 2.

Verify that'the OUADRANT POVER TILT is within its.

Transient Limit yithin;2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the Transient Limit or reduce THERMAL POWER to less than 60%

of THERMAL POWER allovable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months end ieouce the.High Flux Trip Setpoint to < 65.5% of THERMAL POWER allovable.

for the reactor coolant pump combination within the next 4 L'

' hours.

3.

Identify and correct the cause of the out of limit condition-prior to increasing' THERMAL POVER; subsequent

-POWER OPERATION:above 60% of THERMAL POVER allovable for the reactor coolant pump combination may proceed provided that the QUADRANT POWER TILT is verified within its Steady State Limit at least once per hour for.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL POVER.

Vith the OUADRANT POWER TILT determined to exceesi the Transient c.

Limit.but less than the Max 2 mum Limitf-M 1 2. M g due to l

causes other than the misalignment o Rgga_s,algtya,Iegulghg' 'd a

ing or axial power shaping: rods i

^^

1.

Reduce THERMAL POWER to less than 60% of THERMAL POWER allovable for the reactor coolant. pump combination within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpont to $ 65.5%

of THERMAL POVER allovable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POVER; subsequent b

POVER OPERATION above 60% of THERMAL POWER allovable for the reactor coolant pump combination may proceed provided that the QUADRANT POVER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.

DAVIS-BESSE, UNIT 1 3/4 2-10 Amendment No.123 l

J Doditat Nunber 50-346

'~

. Lic:nra Nu2b2r-NPF-3 S; rial. Number 1668 e

' Attachment 1 Page 26

{

P3VER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued)

\\

d.

Vith the OUADRANT POVE Limit 4 M EiT 5 F ( R TILT determined to exceed the Maximum reduce THERMAL POVER to < 15% of RATED 2_ hours.

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SURVEILLANCE REQUIREMENTS i

i' The QUADRART POVER TILT shall be determined to be ri din 4.2.4 limits at least once every 7 days during operation above 1 % of'RITEDC l

THERMAL POWER except when the QUADRANT POVER TILT alarm is inoperable, then the QUADRANT POVER TILT shall be calculated atleast once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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.j DAVIS-BESSE. UNIT 1 3/4 2-11 Amendment No.123

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Dockat Nu:ber 50-346 i

- Lie *nra Nu;ber NPF-3

.S2r,1c1 Nu;b:r 1668 Pige 28 13/4.2.

POLTR DISTRI3UTICN LDiITS I

BASES i

i The specifications of this section provide assurance of. fuel integrity during Condition I (nor=al operation) and II (incidents of moderate frequency) events by: (a) t:aintaining the minimum DNBR in the core 2 1.30 during normal opera-tion and during short term transients, (b) maintaining the peak linear power density s 18.4. kW/f t during normal operation, and (c) maintaining the peak power density less than the limits given in the bases to specification 2.1 during short term transients. In addition, the above criteria must be met in order to meet the assu=ptic s[:ysesL forJhe loss-nfso9hpidents.

3 med 6 tb ContoPERAt:Nc u~its uJearf The power i= balance envelopsj definci 1-Figurae ;.24 and 0.S2fand the insertion' limit curves '

4-"re: 2y pd 2.1--2Jare based on LOCA analyses which have defined the max 2. mum inear~ heat race such that the maximum clad temperature v111'not exceed the Final Acceptance Criteria of 2200*F following a LOCA.

Operation outside of the power imbalance envelope alone does' not con-stitute a situation that vould cause the Final Acceptance Criteria to be ex-ceeded should a LOCA occur. The power imbalance envelope represents the bound-ary of operation limited by the Final Acceptanc iteria_culy if the. control rods are at the insertion limits, as define

'. _ urce 3.1 2 - ud _3.11 )and if 2

the steady-state limit QUADRANT POWER TIL st introduced 'cy application of:

o,yism is -

in & b E OPE @ 4 l.iM M 2Ep m Nuclear uncertainty factors.

a.

b.

Ther=al calibration uncertainty.

c.

Fuel densification effects.

d.

Hot red manuf acturing colerance f actors.

e.

Potential fuel r'od bov effects.

The ACIION statements which permit limited variations from the basic require-ments are accompanied by additional restrictions which ensures that the orig-inal criteria are met.

j The ' definitions of the design li=1t nuclear power peaking factors as used in these specifications are as follows:

I F

Nuclear heat flux het channel factor, is defined as the max 1=a local fuel 9

red linear power density divided by the average fuel rod linear power den-sity, assuming nominal fuel pellet and rod di=ensions.

I DAVII-BESSE, UNIT 1 B 3/4 2-1 Amendnent No..R,30,45 I

L i

..s Docht Nurber 50-346 Lic:nba Nu;ber NPF-3 L

AD0lil0NAl. CHANCES PRfviOUSLT l'5erici Nu'ber.1668 PROPOS(D 6V (fillR

. Attachment 1-Page 29 gg m,q

/gcy ADMINISTRATIVE CONTROLS microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioicdine limit.

MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics, shutdown experience and challenges to tha Pressurizer Power Operated Relief Valve (PORV) and the Pressurizer Code Cafety Valves shall be submitted en a monthly basis to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office, to arrive no later than the 15th of each month following the calendar month covered by the report.

. sy w c / ~-" " hi w v M

?

-n wn

/

/ CORE OPERATING LIMITS REPORT l

J 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

j 3.1.1.3, Moderator Temperature Coefficient 3.1.3.6, Regulating Rod Insertion Limits 3.1.3.7, Rod Program 3.1.3.8, Xenon Reactivity j

3.1.3.9, Axial Power Shaping Rod Insertion Limits 3.2.1, Axial Power Imbalance

\\

3.2.4, Quadrant Power Tilt The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC i

in Topical Report BAW-10122A, Rev.

1,

" Normal Operating 7

Controls", May 1984.

The methodology for Rod Program received L NRC approval in the Safety Evaluation Report dated X X.

The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical limits, core j

thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of

)

' the safety analysis are met.

The CORE OPERATING LIMITS REPORT,

}

including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC

)

Document Control Desk with copies to the Regional Administrator

" and Resident Inspector.

w/

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'% v y_, v, vvw "

DAVIS-BESSE, UNIT 1 6-16 Amendment No. O, 22, H.

106

Dockst Number 50-346 Liccnse'Numbar NPF-3 Serial Number 1668 Page 1 SIGNIFICANT HAZARDS CONSIDERATION Description of Proposed Technical Specification Change The purpuse of this Significant Hazards Consideration is to review proposed changes to the Davis-Besse Nuclear Power Station, Unit Number 1, Technical Specifications. The proposed changes involve relocating cycle-specific core operating limits from the Technical Specifications to a new document entitled Core Operating Limits Report in accordance with NRC Ceneric Letter 88-16.

The Technical Specifications affected and a discussion of the proposed changes is contained in the Technical Description (Attachment 1).

l Significant Hazards Consideration The Nuclear Regulatory Commission has provided standards 10CFR50.92(c) for determining whether a significant hazard exists. A proposed amendment to an Operating License for a facility involves no significant hazards if operation of the facility in accordance with the proposed changes would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated; 2) Create the possibility of a new or.different kind of accident from any accident previously evaluated; or 3) Involve a significant reduction in a margin of safety.

The proposed changes do not involve a significant hazards consideration because the operation of the Davis-Besse Nuclear Power Station in accordance with these changes would:

1.

Not involve a significant increase in the probability or consequences of an accident previously evaluated because there have been no hardware changes or design modifications which would affect the probability or the consequences of an accident.

Dose consequences are unchanged. The relocation of the core operating limits to a new document does not affect the methodology of limit determination and is, therefore, an administrative change only. [10CFR50.92(c)(1)]

2.

Not create the possibility of a new or different kind of accident from any accident previously evaluated because there vill be no hardware changes or design modifications whicL vould create the possibility of a new accident 110CFR50,92(c)(2)]

3.

Not involve a significant reduction in a margin of safety because the j

operating limits vill be determined using the same methodology as in previous core operating limit calculations. [10CFR50.92(c)(3)]

f I

Conclusion Based on the discussion above, it is concluded that the proposed changes do l

not involve a significant hazards consideration.

1 L____

, o

. n.

A i Docket.Numbar 50-346

.j e

License Number NPF-3 Serial Number'1668-

)

g

. Attachment 3' j

4 Page 1 l

j METHODOLOGY

.v l

~

s ROD. PROGRAM

~

a

(

The Core Operating Limits Report has a figure which'provides the Rod Program l

for each cycle. - The Rod Program Figure shows the location of each control

. assembly in-the core and identifies to which rod group the assembly'is assigned. Technical Specification 3.1.3.7 states,."Each control rod assembly (safety, regulating,LAPSR) shall be programmed to' operate in the core location and rod group specified in the CORE OPERATING LIMITS REPORT." The following

-l

~ discussion describes the procedures and methods of analysis used to determine

'the' rods assigned to each group.

l Each control rod assembly in;the core can be electrically connected to any of the eight control rod groups. The only limitation is that each group must have between 4 and 12 rod assemblies. The groups are numbered 1 through 8 and are divided into these categories. Groups 1.through 4 are the safety rods, i

Groups 5 through 7 are the regulating rods, and Group 8 is the axial power shaping rods (APSR's). The following description of the methods and procedures for the analysis addresses the APSR's first and then the regulating rods and finally the safety rods.

j The APSR's are unique in their physical characteristics from the other control' j

rods. They have:a. shorter poison. region and the poison may be of a different I

composition. The' latching mechanism for the.APSR's is.also different from the j

other control rods in~that these rods cannot be unlatched when the control rods are scrammed. The location of the APSR's in the core is symmetric by quadrant. These locations (L12 and N10 in the lower right quadrant) were

)

determined to have a minimal impact on radial power peaking while achieving i

the greatest amount of overall control on core offset (or imbalance). The locations of the APSR's have not been changed since the first cycle and are i

not expected to be changed.

The regulating rods, Groups 5 through 7, are used to control the core power I

level. These rod groups are electrically coupled to be sequentially withdrawn

'(5,6,7), with an overlap of 25% i 5% and are sequentially inserted (7,6,5).

The location of the rods in each of the regulating groups is determined beginning with Group 7.

Group 7 is usually composed of 8 rod assemblies. The location of the Group 7 rod is selected to be symmetrical in the eighth core and have a minimal impact on the radial power distribution.

The locations are also restricted to positions other than those adjacent to the Group 8 rods, if

'l possibic. The Groups 6 and 5 rods are also selected to have eighth core l

I symmetry and have a minimal impact on the radial power distribution.

In addition, the Groups 6 and 5 rods are positioned to ensure an eji :ted rod will I

not violate the safety criteria.

(See BAV-10118A, " Core Calculational l-Techniques And Procedures," J. J. Romano, December, 1979, for a discussion of ejected rod worth analysis).

Design analyses with Groups 5 through 8 determine the limiting rod positions for the operation of each cycle.

These analyses evaluate the operation of the l

j core at various power levels throughout the cycle and show how normal l

operating controls on the rod groups can be set to ensure safe operation. (See i

BAV-10122A, Rev. 1, " Normal Operating Controls," G. E. Hanson, April, 1984, l

)

I

Docket Number 50-346 License Number NPF-3 4

Serial Number 1668 Page 2 for a discussion of rod operation analyses which ensure operating margin with r-espect to power peaking, shutdown reactivity and ejected rod worths).

The safety rods, Groups 1 through 4, are used to ensure sufficient scram reactivity for a safe shutdown of the reactor core. These rod groups are fully withdrawn prior to the core going critical. They are withdrawn one group at a time but not necessarily in order. The core remains shutdown with Groups 1 through 4 out of the core, therefore there is no peaking requirement on the selection of the rod groups for Groups 1 through 4.

However, there are two corAitions which influence the location of the Group 4 rods and the Group 1 rods.

The Group 4 rods may be briefly inserted into the core while the core is in the Startup Mode (2) during physics testing. Thus the location of the Group 4 rod assemblies is chosen such that they are eighth core symmetric. They are also positioned with respect to the location of Groups 5 through 7 to have a vorth of approximately 1.0% ok/k.

The Group 1 rods may be withdrawn any time the core is in either the Hot Standby, Hot Shutdown, or Cold Shutdown Modes by increasing the boron concentration above that required by the shutdown margin with all control rods in.

The reason for withdrawing Group 1 when going through a heat-up te the Startup Pode (2) is to provide an extra margin of safety by having some vorth for a scram should cne be necessary. Thus, the location of the Group 1 rods is chosen to be symmetrical and to have a vorth on the order of 1.0% in reactivity. The results of the Group 1 vorth are also compared to the stuck rod worth such that the reactivity difference can be included in the requirements for the shutdown boron concentration if necessary.

The location of the remaining safety groups, 2 and 3, is selected to be in the remaining locations, but may not be eighth core symmetrical.

If the Group 1 rods are towards the interior, the Group 2 rods vill be more towards the periphery and the Group 3 rods will be more towards the interior.

If the Group 1 rods are towards the periphery, then the location of the Group 2 and 3 rods will be reversed. The Group 2 rods vill be more towards the interior, and the Group 3 rods towards the periphery.

The methods and procedures used to analyze the reactivity and power peaking effects of the control rods are discussed in BAV-10118A and BAV-10122A as noted above.

In order for these analyses to be valid, the location of the rod groups in the core must be the same as those in the Program Figure.

The verification that the electrical connections of the rod groups do indeed correspond to the Rod Program Figure is specified in the Surveillance Requirements of Technical Specification 3.1.3.7.

This surveillance requirement provides the link between the design analyses and core operation.

'4

/*1 Dockat Number 50-346

. Licensa Numbar NPF 'i

, 13......;

z

Serial Number 1668 y

l b

Page 1 SAMPu TOLEDO EDISON L

DAVIS-BESSE UNIT 1 CYCLE 6 i

CORE OPERATING LIMITS REPORT

)

REVISION O i

1.0 Core Operating Limits

]

This Core Operating Limits Report for 09-1 Cycle 6 has been prepared in accordance with the requirements of Tec.hnfcal Specification 6.9.1.7.

The-core operating limits have been developed using the methodology provided i

in the references.

The following cycle-specific core operating limits are incl'aded in this report

1) Moderator temperature coefficient limits

2) Regulating rod insertion limits

3) Rod program group positions

4). Axial power shaping rod insertion limits

5) -Axial power imbalance operating limits and

6) Quadrant power tilt limits.

2.0 References-

1) B&W Fuel Company, Topical Report BAV-10122A, Rev. 1,'" Normal Operating Controls", May 1984

2) B&W Fuel Company, Topical Report BAV-10118A, " Core Calculational j

Techniques and Procedures", December 1979.

3) -Letter from to

, dated

, 1989.

(NRC SER for Rod Program) i

-' ~ - - - - - - - - - - - - - - - - - ' -

~ - - ~ -

R h-[ N.

' DockskNumber'50'-346 L-

- : *?. 1 License Numb 2r NPF-3

(

Serial' Number-1668 SAMPLE' j

' Attachment 4 j

Page 2

~j MODEkaTOR TEMPERATURE. COEFFICIENT LIMITS.

j The moderator temperature coefficient-(MTC)-shall be:

Less positive than 0.9 x 10~4 ok/k/*F vhenever THERMAL POWER is less than a.

95% of' RATED THERMAL POWER,.

b.. Less. positive than 0.0 x 10-4 ok/k/*F: whenever THERMAL' POVER is greater than or equal to-95% of RATED THERMAL POVER, and Equal.to'or less negative than -3.0 x 10-4 ok/k/'F at RATED THERMAL'POVER.

c.

i These limits are referred'to by-Technical Specification 3.1.1.3 I

._u__Ca___m.-___m.. _. _ _ _.

D;ckat Nu:::b:;r 50-346

^

.Licenza Nur.b:r NPF-3 Serial Number 1668 hd(,q El n 3 J

'Page 3 Figure 1 Regulating Group Position Limits, O to 3251 10 EFPD, Four RC Pumps --

Davis-Besse 1, Cycle 6 i

This Figure is referred to by Technical Specification 3.1.3.6 (300,102)

(258,102) 100 - Power Level (270,102)

Cutoff = 100%

(270,92)

SHUTDOWN E

MARGIN g 80 LIMIT (250,80)

]

c.

J E

UNACCEPTABLE G

OPERATION OPERATION 60 S

RESTRICTED

E (170,50)

=

(180,50)

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(128,28.5)

I f 20 ACCEPTABLE OPERATION t (0.0,5.0)

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0-t i

1 0

100 200 300 Rod Index (% Withdrawn) s

)

GR 5 L i

i 0

75 100 GR 6 i

i i

i 0

25 75 100 3R 7 L.

0 25 100

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A Rod Group overlap of 25 5% between sequential withdrawn groups 5 Note 1:

1 and 6, and 6 and 7 shall be maintained.

i m

Dockat Nuib r.'50-346 SAMPLE i

, Licensa Nurbar,NPF-3 l

Serial Number 1668-Page 4 Figure 2 Regulating Group. Position Limits After 325+ 10 EFPD, Four RC pumps, APSRs Withdrawn -- Davis-Besse 1, Cycle 6 This Figure is referred to by Technical Specification 3.1.3.6 4

1 (266,102)

(300,102)

J 100 Power Level l

(270,102) l Cutoff = 100%

SHUT 00WN (270,92)

MARGIN (250,80)

L MIT 80 El

,m a

E$

OPERATION 5

60 UNACCEPTABLE OPERATION f

(176,50)

(180,50) o

{

40 E

E

[

(136,29.5)

$2 20 ACCE/ TABLE OPERATION 8

(0.0,5.0) 0 100 200 300 Rod Index (% Withdrawn)

GR 5 i O

75 100 k

GR 6 L 0

25 75 100 GR 7 i 0

25 100 Note 1:

A Rod Group overlap of 25 5% between sequential withdrawn groups 5 2

and 6, and 6 and 7 shall be maintained.

)

i

Dock 2t Nu bar'50-346 SAMPLE-

.Licenza Nunbar NPF-3 Serial-Number 1668

{

Il j

Page 5 i

i i

Figure 3 Regulating Group Position Limits, j

O to 3251 10 EFFD, Three RC Pumps --

(

Davis-Besse 1, Cycle 6 l

This Figure is referred to by Technical Specification 3.1.3.6 100

}

~

(258,77)

(300,77) 5 80 8

(2 0, 7)

SHUTDOWN l

MARGIN e (270,69.5 )

y LIMIT

)(

(250,60.5) 60

=

EE UNACCEPTABLE OPERATION u

L OPERATION RESTRICTED g

30 (170,38) g (180,38) b 5'

20 (128,21.8)

ACCEPTABLE OPERATION 0

i i

i 0

100 200 300 Rod Index (% Withdrawn)

-GR S '

O 75 100 GR 6 i

i i

i 0

25 75 100 GR 7 '

0 25 100 A Rod Group overlap of 25 5% between sequential withdrawn groups 5 Note 1:

2 and 6, and 6 and 7 shall be maintained.

l i

" ', : Dock t Nuzb;r 50-346' SAMPLE License Number NPF-3 Serial-Number-1668 j

' Attachment 4 6

Page 6 Figure 4. Regulating Group Position Limits After 325 10 EIPD, Three RC Pumps, APSRs Withdrawn' - Davis-Besse 1, Cycle 6 This Figure is referred to by Technical Specification 100 g

80 (266,77)

(300,77)

SHUTDOWN I270'[7) c-MARGIN y

UNACCEPTABLE LIMIT (270,69.5) a OPERATION E

60 250,60.5) 82 OPERATION (176,38)

RESTRICTED 40 5

(180.38) e

.E.

t.

I 20 (136,21.8) c ACCEPTABLE OPERATION 0

O 100 200 300 Rod Index (% Withdrawn)

GR 5 0

75 100 n

GR 6 '

0 43

/s 100 GR 7 '

0 25 100 A Rod Group overlap of 25 5% between sequential withdrawn groups 5 Note 1:

2 and 6, and 6 and 7 shall be maintained.

l

i

+

-O Dock 21 Nunber 50-346 SAMPLE h

(

i License Number NPF 3p g

Serial Numbe'r 1668. '

Figure 5' Control Rod Core Locations

)

.Page 7-and Group Assignments --

i Davis-Besse'1, Cycle 6 This Figure is referred f

x

-N to by Technical Specification 3

3.1.3.7

{

A 3

4 6

4 C

2 5

5 2

D 7

8 7

)

l E

2 5

5 2

1 F

4 8

6 3

6 8

4.

l 1

G 5

1 1

5 i

i

'M w-6 7

3 4

3 7-6

-Y K

15 1

1 5

L 4

8 6

3 6

8 4

M 2

5 5

2

)

N l

7 8

7 0

l 2

5 5

2 P

l.

l l

4 6

4 R

l l

Z 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 Grouc No. of Rods Function 1

4 Safety 2

8 Safety 3

4 Safety X

Group Numoer 4

9 Safety 5

12 Control 6

8 Control 7

8 Control 8

8 APSRs Total 61 h-

5-SAMPLE Dockel Number 50-346

- License Number NPF-3 7

Serial Number 1668

)

- Attachment 4 Page 8 Figure 6 APSR Position Limits, O to 325t 10 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 6 This Figure is referred to by Technical Specification 3.1.3.9 I

i i

RESTRICTED REGION l

100 - (0,102)

(100,102) l 80 l

22 l

LJ l

5l l

a.

s 60 5

PERMISSIBLE cs OPERATING REGION Us

-o 40 Y

8 bb u

I 20 f

r 0

i i

0 10 20 30 40 50 60 70 80 90 100 APSR Position

(*. Witnerawn) a s

-m

1 C-

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SAMPLE

' Dockst Nunb:r 50-346

-License Number NPF-3

.hhh((

]

di

' Serial Number 1668 w El n 5 d

I Page.9 I

_ Figure 7 APSR Position Limits After 1

3251 10 EFPD, Three or Four RC Pumps, APSRs Withdrawn --

Davis-Beste 1, Cycle 6

)

i This Figure is referred to by Technical Specification 3.1.3.9 1

1 100

]$

80 E

c. -

E APSR INSERTION NOT ALLOWED W

60 A:

IN THIS TIME INTERVAL S

^<

Qt Ii 40 a

5 e

E

$[

20 t

sf l

- 1 0

i 0

10 20 30 40 50 60 70 80 90 100 APSR Position (% Withdrawn)

!1 1 -

...", Docket Nurb;r 50-346 SAMPLE License Number NPF-3 h(f Serial Number 1668 si R g

t Page 10' I

gure 8 APSR Position Limits, O to 3251 10 EFPD,-

Three RC Pumps -- Davis-Besse 1, Cycle 6 This Figure is referred to by Technical Specification 3.1.3.9 100 80 RESTRICTED REGIOF a:

L ll (0,77)

(100,77) m.

a E

a:

E 60 S

E ec 40 PERMISSIBLE y

OPERATING REGION i

5S 5

3c.

20 0

O 10 20 30 40 50 60 70 80 90 100 APSR Position (* Witharawn)

p

..l

,L". ' 4 00tkst Nurber 50-346 '

SAMPLE =

License Number <NPF-3.

.=

t.

Serial Number 1668

(-

J Page 11 Figure 9 AXIAL POWER IMBALANCE Limits, Four RC Pumps -- Davis-Besse 1, Cycle 6 i

This Figure is referred to by Technical Specification 3.2.1'

. 110

(-20,102)

(15,102)

~

100

(-25,92)

(15,92) 90 E

(-28,80)6 y-- 80 f20,80) c g-- 70 5

E

.60 w

(-28,50) o j

50_

o (20,50)

"o 40 5

RESTRICTED PERMISSIBLE $ ' -- 30 REGION-OPERATING S.

7 REGION

- 20

$2

~

10 t

t 12 6 t

f f

a f

I I

-50

-40

-3d

-20

-10 0

10 2'O 30 40 50 AXIAL POWER IMBALANCE )

l l..

I

= _ _ _. _ _ _ _

t.

w;o;qg

<. Dc?clut Nunb:r ' 50-346 '

SAMPLE e --

L>

License Number NPF-3 ge m Serial Number 1668 p-s Page 12 Figure 10 AXIAL POWER IMBALANCE Limits, Three RC Pumps'- ' Davis-Besse 1, Cycle 6 This Figure.is referred to by Technical Specification 3.2.1 i

. 110 100 90

~~

(-15,77)

,,(11.25,77)

+

(-18.75,69.5) 70 (11.25,69.5) i

-i

(-21,60.5) 3-- 60 1 (15,60.5) a.

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%-- 30 $

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REGION 5

5 0-- 20 b i3 2

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5 -- 10 j E

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-30

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-10 0

10 20 30 40 50 1

AXIAL POWER IMBALANCE (%)

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