ML19317E859

Amends 34,34 & 31 to Licenses DPR-38,DPR-47 & FPR-55, Respectively,Establishing Operating Limits for Unit 3 Cycle 2 Operation Based on Acceptable ECCS Model & Terminating Operating Restrictions Imposed 741227 on Unit 3

ML19317E859
Person / Time
Site:	Oconee
Issue date:	10/22/1976
From:	Goller K Office of Nuclear Reactor Regulation
To:
Shared Package
ML19317E858	List: ML19317E857 ML19317E859 ML19317E860
References
NUDOCS 8001031032
Download: ML19317E859 (34)

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Dttri POWER COMPANY

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DOCKET'N0. 50-269 t

a OCONEE NUCLEAR STATION,tNIT 1 AftliDWENT TO FACILITY OPERATIUG LICENSE Arsendnent No.3//

L1 cense No. DPP.-38 r

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The Huclear Regulatory Co:mlssion (the Comeinica) has found thatt A.

The application for amendment by Duke Power CompanyYthe _

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licensee) dated July 21, 1976, as sc;ple?.ented August 20, October 7, October 1 S October 2 Q and October 2 0,1976, comply with the standards and require =ents of the Atoeic Energy Act of 1954, as anended (the Act), and the Cesatssion's rules and regulations set forth in 10 CFR Chapter 13 -

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The facility wfil operate in conformity wf th the appitcation, s

the provisions of the Act, and the rules and_ regulations of the Cormission; j

3, C.; There is reasonable assurance (f) that the activities authorized by this acendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in coop 11ance with the Connission's regulations; l

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The issuance of this anced:ent will not be inir41 cal to the

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cornen defense and seccrity o'r to the health and safety of the public; and t

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The issuance of this amendzent is in accordance with 10 CFR Part 51 of the Ccruission's regulations and all applicable requirements ~ ', x have been satisfied.

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Accordingly,' the license is ameMed by chanFes to the Technical ~

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Spectfications as indicated in the attachment to this. license

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asentment, 1

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This license seendment is effective as of the 'dste of its issuance.

FOP. THE NUCLEAR REGULATORY C0FmISSION original signed br' Karl R. Go1Hr. Assistant Director for e

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Specifications Date of Issuance:

OCT 2 21976 '

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OCONEE NUCLEAR STATION ' UNIT 2:,.'

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Amendment No. M

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License No. DPR-47

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A.~n The application for issendment by,Duks Power Compaqr.(tbeitVjM ' t e

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-:- N licensee) dated July 21, 1976, as supplemented August 20, V, -

October 7, October 1 g, October 2 0, and OctEber2:0,;1976,"

cocply with the standards and requirements of the Atomic:.

Energy Act of.1954, as amended (the Act).'.end the Coussission's

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rules and regolations set forth in 10 CFR Chapter II B.

The facility will operate. in confensity with the application.,-

the provisions of the Act, and the rules.and regulations of,

the Commission;

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There is reasonable assurance'(1) that the activities authorized by this amendsent can be conducted without endangerine the ';

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will be conducted in compliance with the Commission's regulations;[ _

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1 FOR THE NUCLEAR REGULATORY COMMISSION l

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DOCKET No. 50-237 OCONEE NUCLEAR STATION, UNIT 3

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AHFHDMENT TO FACILITY OPERATING LICENSE Amendment No. J/

License No. CPR-55 s

1. The Nuclear Regulatory

• Commission (the Ccomission) has found that:

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The' application for amendment by Duke Power Company (the-A.

licensee) dated July 21,1976, as supplecented August 20, October 7. Oct6ber 1 S, October 2 0, and October 2 D,1976, conply with the standards and requirments of the Atomic

' Energy Act of 1954, as amended (the A:t), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

Thh facility will' operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized' by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in coup 11ance with the Cor. mission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to. the health and safety of the public; and

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The issuance of this annement is in accordance with 10 CFP. Part_

51 of the Commission's. regulations and all applicable requirments

. have been satisfied.

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-u ATTACHMENTTOLICENSEAMENDMENTS AMENDMENT NO. 34 TO DPR-38 AMENDMENT NO. 34 TO DPR-47 AMENDMENT NO. 31 TO DPR-55 00CTITS NOS. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:

Remove the following pages:

2.1-3c 3.5-8 3.1-17 f

2.1 -3d 3.5-9 2.1 -6 3.5-10 i

2.1 -9 3.5-11 2.1-12 3.5-16 2.3-2 3.5-16a 2.3-3 3.5-17 2.3-7 3.5-20 2.3-10 3.5-23 2.3-13 3.5-24 3.5-7 4.1-9 Insert identically numbered pages, as above.

Add pages:

3.5-20a 3.5-20b 3.5-23a 3.5-23b Delete pages:

3.17-1 3.17-2

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_ Benes - Unit 3 2

. N safety limits preaanted for ocones Unit 3 have been generated using B&W-2 crieie=1 heat. flux corr =1stion(l) and the teactor Coolant System flow rate of 107.6 percaat of the design flow (131.32 x 106 1he/hr for four-pump oper=tian).

The flow rate utiliaad is conservative compared to the actual measured flow rate.(2)

To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB).

At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure.

I Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, tempera i

and pressure can be related to DNB through the use of the BAW-2 correlation I

The RAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the marain to DNB.

The miniana value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30.

A DNBR of 1.30 corresponds to a 95 percent probability at a 95 percent confi-dance level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.

The difference in these two pressures is nominally 45 pai; however, only a 30 psi drop was j

assumed in reducing the pressure trip serpoints to correspond to the elevated location where the pressure is actually measured.

l The curve presented in Figure 2.1-1C represents the conditions at which a J

minisum DNBR of 1.30 is predicted for.the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 141.3 x 106 lbs/hr.). This curve is based on the following I

nuclear power peaking factors with potential fuel densification and fuel rod bowing eHects

= 2.67; F

= 1.78; F,

= 1.50.

The design peaking AR combination results in a more conservative DNBR than any other power shape that exists during normal operation.

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, The enryms of Figura 2.1-2C are based on the more restrictive of two thermal 14=ien and i=e1=da the effects of poemmetal fuel d==*He=*4= and fuel rad bowing.

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The 1.30 Jamt 1fmit produced by a - peaking factor of F I = 2.67 or thecombinationoftheradialpeak,axialpeakandpositionoftheaxial peak that yields no less thma a'1.3015B1.

2.1-3e Amendments Nos. 34, 34 & 31 f

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The combination cf radial and' axial peak that causes central fuel melting 2.

The limit is 20.15 kw/ft for Unit 3.

at the hot spot.

Power p==1 ring is not a directly observable quantity, and, therefore, limits beve been established on the bases of the reactor power 4=u--a produced by the power peaking.

The specified flow rates for Curves -1,- 2 and 3 of Figure 2.1-2C correspond to the exnected minimum flow rates with four pumps, three pumps and one pump in each loop, respectively.

l The curve of Figure 2.1-1C is the most restrictive of all possible reactor coolant M --wi--

thermal power cambia =t4 = = ahown in Figure 2.1-3C.

The maximum thermal power for three-pump operation is 86.4 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow z 1.07 =

79.9 percent power plus the==v4=u= calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a j

afmilmr mannar.

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For each curve of Figure 2.1-3C a pressure-temperature point above and to the left of the curve would result in a DNER greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor The 1.30 DNBR curve for four-pump operation is more 7

coolant pump. situation.

restrictive than any other reactor coolant pump situation because any pressure /

temperature point above and to the lef t of the four-pump curve will be above and to the left of the other curves.

References (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized I

Water, BAW-10000, March 1970.

(2) Oconee 3, Cycic 2 - Reload Report - BAW-1432, June 1976.

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i 2.1-3d Amendments Nos. 34, 34 & 31 a-e.

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Reactor Outlet Temperature, F l

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CORE PRDIETIDit SAFETY LIMITS UltIT 3 OCONEE NUCLEAR STATION

'2.1-4 Figure 2.1-1C Anerdnents Nos. 34, 34 3 31 l

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Acceptable 4-Pump i

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(30. 8. J6.4) a 80

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2. 3 & 4 Pump 43, 46.9)

Operation

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60

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-20 20 40 60 Reactor Power Imbalance, %

Curve Reactor Coolant Flow (Ib/h) 1 141.3 x 106 2

105.6 = 106 3

69.3-= 108 1

CDRE Fxulu,IIDN SAFETT LIMITS UNIT 3 2.1-9 OCONEE NUCLEAR STATION Figure 2.1-2C Amendnents Nos. 34, 34 & 31 f

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560 580 600 620 640 660 Reactor Outlet Temperature, F Reactor coolant flow Curve (1bs/h)

Power Pumps operating (type of limit) 1 141.3 x 106 (100%)

112%

Four pumps (DNBR limit) 2 105.6 x 106 (74. 7%)

86.4%

Three pumps (DNBR limit) 3 69.3 x 106 (49.0%)

58.9%

One pump in each loop (quality limit)

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PROTECTION SAFETT W e MEE NUCE. EAR S 2.1-12 F1 ure 2.1-3C 9

f Amendments Mos. 34, 34 lk 31

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L, 1 haring normal plant operation with all reactor coolant ptasp] operating.

, " reactor trip is initiated when the reactor power level reaches 105.5% of rated power. F ?ing to this the possible variation in trip setpoints due to calibration a.;d instrument errors, the mari== actual power at which a trip would be actuated could be 112%, which is more conservative than the value used in the safety analysis. (4)

Overpower Trip Based on Flow and Imbalance i

The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accotasodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has demenstrated that the specified power-to-flow ratio is adequate to prevent a DNBR of less than 1.3 should a

. low flow condition exist due to any electrical malfunction.

1 The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power-to-flow ratio provides overpower DNB pro-tection for all modes of pump operation.

For every flow rate there is a maxi-mum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situtations of Table 2.3-1A are as follows:

1.

Trip would occur when four reactor coolact pumps are operating if power is 105.5% and reactor flow rate is 100%, or flow rate is 94.8% and power leve'l is 100%.

[

2.

Trip would occur when three reactor coolant pumps are operating if power is 78.8% and reactor flow rate is 74.7% or flow rate is 71.1% and power level is 75%.

3.

Trip would occur when two reactor coolant. pumps are operating in a single loop if power is 51.7% and the operating loop flow rate is 54.5% or flow rate is 48.5% and power level is 46%.

4.

Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 51.7% and reactor flow rate is 49.0% or flow rate is 46.4% and the power level is 49%.

The flux-co-flow ratios account for the maximum calibration l

pad instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

i For safety calculations the maximum calibration and instrumentation errors for the power level trip were used.

j The power-imbainee boundaries are established in order to yeuvent reactor M. thermal limits from being exceeded. These thermal limits are either power pa=Hng kw/ft limits or DERE limits. The reactor power M-1=nca (power in the top half of cora minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio much that the boundarias of N.

l-Figure 2.3-2A - Unit 1 are produced. The power-to-flow ratio reduces the power]

2.3-2C - Unit 3

,0399 D gD m p ' QI 2.3-2n - Unit 2

'cun a A 2.3-2 1

• c~.i. eats Nos. 34, 34 1 31

.w---

~

sip and associated reactor power / reactor power-imbalance boundaries

. _-55%-Unit 1 for a II flow reduction.

g 1.-

za - Dnit 2 U

5 - Unit 3 U__

st 1, the power-to-flow reduction ratio is 0.949, and for Unita 2 and 3,

_ j.,r r -

- -to-flow reduction factor is 0.961 during single loop operation.

mitors

- monitors prevent the minimum core DNBR from decreasing below 1.3 by W-g the reactor due to the loss of reactor coolant pump (s). The circuitry e-

- mg pump operational status provides redundant trip protectica for DMB C[

ging the reactor on a signal diverse from that of the power-to-flov

%e pump monitors also restrict the power level for the number of p operation.

v" Coolant System Pressure

_sa startup accident from low power or a slow rod withdrawal from high V'

he system high pressure set point is reached before the nuclear over-

~ -:::::-ip set point.

y The trip setting limit shown in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2 2.3-1C - Unit 3 6:8:' gj reactor coolant system pressure (2355 psig) has been established-to the system pressure below the eafety limit (2750 psig) for any (1)

-::ransient.

W p pressure (1800) psig and variable low pressure (11.14 T

-4706) trip

~

(1800) psig ut (10.79 T

-4539)

(1800) psig ut (10.79 T

-4539) p shown in Figure 2.3-1A have been established to maintalk the DNB

{

2.3-1C g: eater than or equal to 1.3 for those design accidents that result in s f,,scre reduction (2,3) go the calibration and instrument.ction errors the safety analysis used a

• • 1. abu low reactor coolant system pressure trip value of (11.14 T

-4746)

(10.79 T,

-4579)

(10.79 T

-4579)

{

out M nne Outlet Temperature i

.fhe high reactor coohant outlet temperature trip setting limit (619 F) shown Fisura 2.3-1A has been established to prevent excessive core coolant 2.3-1B 2.3-1C

,,,,eratures a the o,emag -e. one to ca uhratism and a. u u.a arters the safety analysis used a trip set point af 62D E.*

anduiam Pressure

~

~

p high rametar building pressure trip setting limit (47sig) provides p,,ggive assurance that a reactor trip wilroccur in the.unlikely ennt of

, 3,,,.of-coolant accident, even in the absence of a low reactor coolant pas pressure trip.

~

2.3-3

{

Amendments Mosr 34,. 34 & 31 l

4

~ _ _ _

g level trip and associated reactor power / reactor power-imbalance boundaries by 1.0551-Unit 1 for a 11 flow reduction.

1.07% - Tinit 2 1.07%"- Unit 3 For Unit 1, the power-to-flow reduction ratio is 0.949, and for Unita 2 and 3, the power-to-flow reduction factor is 0.961 during single loop operation.

Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss af reactor coolant pump (c). The circuitry monitoring pump operational status ptovides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio. The pump monitors also restrict the power level for the number of pumps in ope ation.

Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdrawal from high i

power, the system high pressure set point is reached before the nuclear over-power trip set point. The trip setting limit shewn in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2

. 2.3-1C - Unit 3 J

for high reactor coolant system pressure (2355 psig) baiFbeen established-to maintain the system pressure below the safety limit (2750 psig) for any design transient. (1)

The low pressure (1800) psig and variable low pressure (11.14 T

-4706) trip

~

(1800) psig (10.79 T "*"-4539)

(1800) psig (10.79 T " -4539)

{

setpointsshowninFigure2.3-1AhavebeenestablishedtomaintaEf1 the DNB 2.3-1B 2.3-1C ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2,3)

Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 T,,g -4746)

(10.79 T

-4579)

(10.79 T,oug -4579)

{

e

,g Coolant Outlet Temperature The high reactor coolant outlet temperatur p y setU,ng limit (619 F) shown in Figure 2.3-1A has been established tc p. p!M. ;urcessive core coolant 2.3-1B 2.3-1C Due to ca7 n, rat 164 and instrimentation temperatures in the operating range.

errors, the safety analysis used a trip sat poi.2 of 620 f.,

• Emactor hi1di== Pressure 1

.3he high reactor building pressure trip setting limit (4 psig) prowli.es positive assurance that a reactor trip 'tiill' occur in e.e.unlikely event of a loss-of-coolant accident, even in the absence of a low rector coolant system pressure trip.

1 knendments 190s734, 34 & 31

e l

6 2400 F = 2355 psig T = 619F 1

2300 2200 i

3 E.

8 E

Acceptable f

2100 Operation j

m c2

... 8 u

E og 2000 a

o y

\\

\\

1, o R

Unacceptable o-Operation 5

1900 t

a 1

P = 1800 psig 1800 (587.5)

I 1

I t

540 560 580 600 620 640 Reactor Outlet Temperature F

_ s_ __

ALL0hlABLE SETPOINTS

~

2.3-7 UNIT 3 l

OCONEE NUCLEAR STATION Figure 2.3-1C Amendments Mos. 34, 34 & 31 m

f'

,)e a

/

Power level, I

.J.g

(-11,. 107)

(18 _107)

I Pour Pump Setpoint

- - 100

+

s

(-28, 93) h

~

6(30, 90)

• 1hree Pump Setpoints Rn 11, 79.9 ) (18, 79.9

(-28, 65.9)

<(30, 62.9)

• Two Pump

' ~

Setpoints

-11, 52. 4)

(18, 52.4)

?

(-28, 38.4)

- 40

>(30, 35.4)

C 20 o

o Y

Y a

e a

n

[

m" d

u O

60

-40

-20 0

20 t-1 i

40 60 i

Power Imbelance, 2

^

9 PRDTECTIVE SYSTEM MAXI M ALLOWABLE SETPOINTS 2.3-10

1. 3 I

OCONEE NUCLEAR STATION Figure 2.3-2C

[

Amendments Nos. 34, 34 & 31

-J 4

^

[.

E Table 1.3-1C Unit 3 Reactor Protective Syy, ten _Tr_fy Ey,tt,f_ng I.fmite i

(

Two Reactor One Reactor Four Ac.orter Three Reactor Coulent Pawape Coolant Puer l

Conlant Pumpe Coolant Pi+9e Ope rat f ag in 4 Operating te 3

Operating Oper.iting Single Taop F.ach loop (Oper.stlug Power (Operating Power (Operat ing Power (Operating Shut duwie RFS, fermi nj, 1002 R.ited) f

-75%Ratedl

-4{I_Rjted)_

-492 Rated).

Bypase._

1.

leuclear Power nice 105.5 105.5 105.5 105.5 5.0III (t Rated)

2. *lhaclear Fuwer nine Essed 1.07 times flow I.07 times flow 0.961 times flow I.07 times flow Sypaamed en Flow (2) and inhalmsre, etnue reduction afnus reduction minus reduction afnus reduction (I RAled) 6 due to Imbalance due to Imbalance due to Imbalance due to fabalance 3.

Ilocic4r 1%wr neue it.vicd NA 1A SSI (5) (6) 553 Bypassed ene Pump thetture, [8 Rate.1) l 4 Nigle Reactor coolant 2355 2355 2355 2355 1720 Systre Pressure, pels, Men.

II

.d 5.

Imv Reactor omlaat l 1800 1800 1800 1800 sypassed Y

Systre recenuree puls. Hin.

6.

arlebte law Reactor (10.79 7""" 4539)III (10.79 7'"" 4539)II}

(10.79 7""" 4539)III (10.79 T -4539)III sypassed tant yeten Pressure

• • I alge Min.

g l

8 l

e F.

Reactor foolant Teg.

619 619 619 (6) l

!,l.

F.,Plese

, 619 619 i-Nigh Reetter hatidies 'I

'll 3.

I i

4 4

'4 e

'4 4

Pressure, pels, e.-

i

..L.j

..l 7

Q

_ 'j l

()) f d 18 m degrees sheenheit (*F).

Q (5) Desctor power level trip set point produced e

by pump contact monitor reset to 55.02.

.l,'

l (2) Reacter Coolene gheten Flow. 2.

C Q

(3) Adminlettatively enntrolled reduction set (6) Specification 3.1.8 applies. Trip one of the f

only during reacter ebutdown, tuo protection channele receiving outlet n

en temperature information from sensore in the (4) Automatitetty set sleen other segments of idle loop.

l i

g the hr5 are bypsesed.

cc::~3 o

===>

N c%

i N

f r

e - - - --

j s.

t 3.1.7 Moderator Temperature Coefficient of Reactivity Specification The moderator temperature coefficient shall not be positive at power levels above 95 percent of rated power.

Bases A non-positive moderator coefficient at power levels above 95% of rated power is specified such that the maximum clad temperatures will not exceed the Final Ac-ceptance Criteria based on LOCA analyses. Below 95% of rated power the Final Acceptance Criteria will not be exceeded with a positive moderator temperature coefficient of +0.9 x 10-4 Ak/k/ F. corrected to 95% rated power. All other ac-cident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0.9 x 10-4 Ak/k/ F.

The moderator coefficient is expected to be zero or negative prior to completion of startup tests.

When the hot zero power value is corrected to obtain the hot full power value, the following corrections will be applied.

A.

Uncertainty in isothermal measurement The measured moderator temperature coefficient will contain uncertainty on the account of the following:

.1.

+0.2*F in the AT of the base and perturbed conditions.

2.

Uncertainty in the reactivity measurement of +0.1 x id-4 Ak/k.

Proper corrections will be added for the above conditions to result in a conservative moderator coefficient.

i B.

Doppler coefficient at hot zero power During the isothermal moderator coefficient measurement at hot zero power, the fuel temperature will increase by the same amount as the moderator. The measured temperature coefficient must be increased by 0.16 x 10-4(Ak/k)/*F to obtain a pure moderator temperature coefficient.

a Modtrator temperature change I

L The hot zero power measurement must be reduced by.09 x 10-4 (Ak/k)/*F.

This corrects for the difference in water temperature at zero power (532*F) and 15% power (580*F) and for the increased fuel temperature effects at 15% power. Above this power, the average moderator temperature remains 580*F.

However, the co-efficient, s., must also be adjusted for the interaction of an m

average moderator temperature with increased fuel temperatures.

This correction is.001 x 10-4 Am /A% power.

It adjusts the 15%

m to the moderator coefficient at any power level above 15%

power a s

)

4 power. For exam

(.001 x 10-4) (ple, to correct to 100% power, u85%), which is

.085 x 10-4 m. m is adjusted by

\\

A I

m

?

3.1-17 Amendments Nos. 34, 34 & 31 1

L

-a

.~

~

If within one (,1) bour of determinati g.

combining the worth. of the danpar=hla ro utdown margin exists rods, the teactor.shall be brought to the hotwith each,of t 1

until this margin is established.

standby cond4*4 =__

h.

Following the determination'of an inope

.be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and ex~arcised wee problem is solved.

ntil tha rod i.

If a control rod in da regulating or declared inoperabia, power shall be redsafety rod groups is

~

the thermal power allowable for th uced to 60 percent of bination.

e reactor coolant pump com-j.

is declared inoperableIf a control rod in the regulating o}

operation above 60 percent of ratedaxia power may continue prov,ided the rods in th

\\

such that the rod that was declared inoper ble group are po within allowable group average positio e is maintained

}

a 3.5.2.2.a and the withdrawal limits of Sn limits of Specificationj 3.5.2.3. '_.__ The worths af_. single pecification 3.5.2.5.c.

4 are limited by.

inserted controLroda during critic li

. control rod pos.the restrictions of specification 3.1.M 1

Quadrant Power Tiltition limits defined in Specificati a

ty..

3.5.2.4 on 3.5.2.5.

a.

tilt exceeds +3.41% Unit 1Except for physics tests, if t

, either the quadrant power tilt shallx 3.41% Unit 2 3.41% Unit 3 be reduced to less than +3.41% Unit I 3.41% Unit 2 within two hours or the

[

following actions shall be taken:3.41% Unit 3

{

(1) If four reactor coolant pumps thermal power shall be reduced belare in operation, the allowable by 2% of full power for each 1% tilt (as identif ow the power level cutoff

.5) and further reduced j

l in excess of 3.41% Unit 1.

3.41% Unit 2 (2) If less than four reactor coola 3.41% Unit 3 l shall be radneed by 2% of full power fallouab coolant pump combination or each*1% tilt.

ga<

3.5-7

• 2A

_y Amendnents Nos.

\\

34, 34 & 31 t

y.

t (3) Except as provided in specification 3.5.2.4.b. the reactor j'

shall be brought to the hot shutdown condition within four hours if the quadrant power tilt la mot adne=d to laas than J

3.41% Unit I within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.41% Unit 2 3.41% Unit 3-I b.

If the quadrant tilt exceeds +3.41% Unit 1 and there is simultaneous 3.41% Unit 2 3.41% Unit 3 indication of a misaligned control rod per Specification 3.5.2.2,

[

j reactor operation may continua provided power la reduced to 60%

j of the thermal power allowable for the reactor coolant pump combination.

Except for physics test, if quadrant tilt exceeds 9.44% Unit 1, c.

9.44%. Unit 2 9.44% Unit 3 l-a controlled shutdown shall be initiated immediately, and the reactor shall be brought to the hot shutdown condition within four hours.

d.

Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4.a(3) or 3.5.2.4.c above, subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the power I

range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each 1 percent tilt for the==v4==

tilt observed prior to shutdown.

t Quadrant power tilt shall be monitored on a =4n4= m frequency e.

of once every two hours during power operation above 15 percent of rated power.

t 2.5.2.5 Control Rod Positions 1

Technical Specification 3.1.3.5 does not prohibit the exercising a.

of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

b.

Operating rod group overlap shall be 25% 1 5% between two sequential groups, except for physics tests.

Escept for physica **=es or exeref =4== control roda. the enneval c.

F rod withdr== 1 11=4em are specified on Figures 3.5.2-1A1 and I '

3.5.2-1A2, (Unit 1), 3.5.2-1B1, 3.5.2-152 and 3.5.221B3 (Unit 2),

and 3.5.2-lC1, 3.5.2-1C2, and 3.5.2-lC3 (Unit 3) for four pump operation and on Figures 3.5.2-2A1. 3.5.2-2A2 (Unit 1). 3.5.2-231.g 3.5.2-2B2 and 3.5.2-2R3 (Unit 2), and 3.5.2-2C1, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) for three or two ymmy

]

W t

Amendments Mos. 34, 34 4 31

._._e, mo mr..

a.

m

^

-h

i

..).-

/

operation.

If the control rod position limits are exceeded, corrective measures shall be taken lamediately to

. achieve an acceptabla control rod. position.

Acceptable contral rod position shall then be attained within tan hours.

The mini== shutdown margin required by Specification 3.5.2.1 shall be maintained at all times, d.

Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-161, 3.5.2-1A2 (Unit 1), 3.5.2-1B1, 3.5.2-152, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, 3.5.2-IC3 (Unit 3), unless the following requirements are met.

(1)

The zenon reactivity shall be within 10 percent of the value for operation at steady-state rated power.

i (2)

The xenon reactivity shall be asymptotically approaching the l

value for operation at the power level cutoff.

3.5.2.6 Reactor power imbalance shall be monitored on a frequency not to I

Except for physics tests, imbalance shall be maintained w envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B1,' 3.5.2-352, 3.5.2-353, 3.5.2-3C1, 3.5.2-3C2, and.3.5.2-3C3.

If the imbalance is not within the envelope defined by these figures, corrective measures (

imbalance is not achieved within two hours, reactor p

~

reduced-until imbalance limits are met.

3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the manager.

s

~

o

' 3-9 Amendments Nos. 34, 34 & 31

% n.M*s-4 m r.

= w x -

Bases The power-imbalance envelope' defined in Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-351, 3.5.2-3B2, 3.5.2-353, 3.5.2-3C1, 3.5.2-3C2 and 3.5.2-3C3 is l

based on LOCA analyses which have defined the maximum linear hcat rate (Sea Figure 3.5.2-4) such that the maximum clad temperature will not exceed the Final Acceptance Criteria. Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary. Operation in a situation that vo.1d cause the Final Acceptance Criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.** Conservatism is introduced by application of:

a.

Nuclear uncertainty factors b.

Thermal calibration

~~

c.

Fuel densification effects d.

Hot rod manufacturing tolerance factors

~-

I, -

The 25% + 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.

Control rods are arranged in groups or banks defined as follows:

Group Function 1

Safety 6

2 Safety

)

3 Safety 4

Safety 5

Regulating 6

l Regulating

~7 Xenon transient override 8

APSR (axial power shaping bank)

The rod position limits are based on the most limiting of the following three criteria:

ECCS power peaking, shutdown margin, and potential ejected rod worth. Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consis-tent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1). The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.5% Ak/k (Unit 1) or 0.65% Ak/k (Units 2 and 3) at rated power. These valmas have been shown to be safe by_the. safety analysis 12,3,4) of the hypothetical rod ajectime accident. A single inserted control red worth of 1.0% Ak/k is allowed by the rod positions limits at hot zero power.-

A single inserted control rod worth of 1.0% Ak/k at. beginnin'g-of-life, hot

(

sero power would result in a lower transient peak thermal powr and, there-1 l

fore, less severe environmental consequences than a 0.5% Ak/e (Unit 1) or

- 0.65E Ak/k (Units 2 and 3) ejected rod worth at rated pouer.

• • Actual operating limits dep-and on whether or not incere~or excora detectors are used and their respective instrument and calibration errors. The method used to define the operating limits is defined in plant operating procedures.

t j

Control rod groups are withdrawn in sequence beginning with Group 1.

Groups 5, 6, and 7 are overlapped 25 percent. The normal position at power is for Groups 6 and 7 to be partially inserted.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established with consideration of potential effects of rod bowing,

and fuel densification to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 5.10% for Unit 1.

The limits shown in Specification 3.5.2.4 5.10% for Unit 2 l

5.10% for Unit 3 l

are measurement system independent. The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the station operating procedures.

The quadrant tilt and avial imbalance monitoring in. Specification 3.5.2.4 sad 3.5.2.6, respectively, normally will be performed in the process I

computer.

The two-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.

Allowance is provided for withdrawal limits and reactor power imbalance limits'to be exceeded for a period of two hours without specification violation.

Acceptable rod positions and imbalance must be achieved within

^

the two-hour time period or appropriate action such as a reduction of power taken.

Operating restrictions are included in Technical Specification 3.5.2.5d l

l to prevent excessive power peaking by transient xenon. The xenon reactivity must be beyond the "undershoot" region and ssymptotically approaching its equilibrium value at the power level cutoff.

_ REFERENCES bSAR,Section3.2.2.1.2 2FSAR, Section 14.2.2.2 FSAR, SUPPIJMENI 9 h JUEL 3DISIFTCATT M IEPORT BAU-1409 (UNIT 1)

~

".ms-1400 (UEIT M 1

i

=

3*hll kendnents Nos. 34, 34 & 31

1 I

\\

i

)

i 170, 102O O202.5, 102 100

• 1 90 170, 91

)

< 202.5, 91

~

161, 85 Restricted 80 Restricted Region hgie a

E 70 25.8, 65 300, 64 O

1 9, 62 60 I

y i

l 50 I

Permissible 40 Operating f

Region i

30 1

20 10 0, 0 0

e i

8 i

i i

e i

e i

i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, % withdrawn i

0 25 50 75 100 0

25 50 75 100 I

I t

i 1

1 l

e 3

Group 5 Group 7 0

25 50 75 100 a

l l

l t

Group 6

. ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROP l'

- 0 TO 115.(110) EFPD

~

UNIT 3

1. NE"# N M N 3.5-16 Figure 3.5.2-1C1 Amendments Nos. 34, 34 1 31

.-,..,.c.,-___

~ _ _ -.,.,

)

• l l.

)-

l 122,102 Operation in thin 170,10g

'09.4, 102 100 6

kenion in Not 90 Allowed 170, 91 j

g 209,4, 91 P W" level

""L"II g(i - Sissitdown 215 H5 M.srgin l

E Restricted 1.im i t Restricted 70 -

se Region Region 129, 62 I'#

.o 225.a. e.3 9-C

~

25, 50 i

h 40 Permissibac Operating, Hegion 30 -

l 0, 15 10-h

eb a

i e

r t

e t

g 0

2i 40 60 60 100 120 140 160 180 200 220 240 260 280 300 Rod Index. I withdrawn 0

25 50 75 100 0

25 50 75 300 t

i e

f a

a I

f n

i Group 5 Croup 7 0

25 50 75 100 i

f f

f I

Group 6 c

aco pos1130u LIMITS TDit FOUR

~

PUHP OPERA ION FRG4115 Q 10)

~

CrPD TO F26 Q 10) EFPD trail 3 OCONEE 9eucLEAR STAT 80N

~

- h 5-14 Figure 3.5.2-102 Amendments Nos. 34, 34 & 31 f> '

• a

~-

~

~ ~ - -

~

. ~..

i jg_

Operation in this I'd 102 255 102 RcKien is Not j

Allowed Power 1.evel 90 5

Cutoff 240, 91 80 Rest ricted E

Stiu tdown y

S Harr,in N

60 96, 67 225.8. 68 -

o

63. 50 50 g

40 Permissible

^-

. Operating 30 Region j

20 0,2 30 -

O', O 0,

n.

i e

i i

e u

20 40 60 80 300 120 140 160 180 200 220 240 260 280 300 R d Index, % withdrawn 0

L U

I0 75 100 0

25 50 75 100 8

t 1

g Croup 5.

3 Group 7 9

0 25 5,0 75 10, 0 1..-

e f

Croup 6

~

ItDD P051110ft 11MITS 7VR FDL ID*F (TLRA1200 AFIIR 226

(+. 10) U PD tmIT 3

~

9 OCONEE-NUCLEA m

~

3.5-17 iigure 3.5.2-1C3 I

Amendments 1tos. 34 M & 31 p=-

~L

~

~

o i

i Restricted Region 11 102 1 4,102 2,102 Rest.s.ted 100 g

- for 2 and 3 Pump Region for Operation 3 Pump 4,4 90 operation O

8 26,83 300,82 80 4U 129,79 u

4, 6

50 Permissible Operating Region 40 R

08 30 20 s'.

k 10 o.

I t

f I

I I

I e

I I

I t

e 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, % withdrawn 0

25 50 75 100 0

25 50 75 100 a

a e

t i

n l

l g

I Group 5 Group 7 1

0 25 50 75 100 I

e i

i i

Group 6 l

l ROD POSITION LIMITS FDR h AND THREE-PUMP OPERATION FROM 0 TO 115 (+ 10) EFPD UNIT 3 l

3.5-20 G MEAR STATION

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Figure 3.5.2-2C1 Anandmants Nos. 34. 34 & 31

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Operation in this 122 102 46,102 23 8 02 Restricted 4

Region for g100 Region is W t 3 Pump Allowed O eration 3

P "5

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226,8;-

300.82'

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I 30 Shutdown 129,79 g,,gg I

u Limit 4*

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=

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• 20 M 15 k10 l

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20 40 60 80 100 120 140 160 180 200 220 2/.0 260 280 300 j

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Rod Index, % withdrawn 75 100

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25 50 75 300 0

2,5 50, i

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i Croup 7 Croup 5 O

25 50 75 100 t

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Group 6 RCD POSITION LIMITS FOR TWD-

1. m TilREE-PIC P CrERATION FR' 115 (* IO) 70 226 Q 10) IO

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UEIT 3.

, OCONEh. NUC1. EAR STAT 10N

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3, go, rimwe 3.5.2-2C2 Amendments Nos. 34, 34 1 31

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100 Operation in this Restricted Region S

Region la Hot for 3 Pump g

Allowed Operation 3

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80 Shutdown Marsin 20,85

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k 70 -

P, 60 Permissible Operating-Region s

l 50 63.5

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3.

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p 30 20

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15 go Restricted for 2 and 3 Pump Operation

0. 0, l,

0 20 40 1 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, % withdrawn 0

25 50 75 100 0

25 50 75 10 i

i Croup 5 Croup 7 0

25 50 75 100 1

1 I

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Group 6

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ROD POSITI0ff 1.1MYTS FO':

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~'~. Af;D THREE-?Uf9 OPERATie

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_ AFTER 226 (+ 10) EFPD I

' -.UfJIT 3 3.5-20b..

' ~ ~

OCONEE NUCLEAR STATIC 1

i P._ Figure 3.5.2-2C3

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- t 1

o i

l 1

Power,.E of 2568 Hun Restr$ted Region

- -110

-7.96,102 16.83,102

-100 j

15.22,M i )

90

-7.78,91

-9.67,8 320.12,86

-80

-70 1

23.64.64(

-60 l

Permissible

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-50

_ Region 40

-30

'6 a

-20 l

1

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f 50 -40 -20

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10 20 30 40 50-J Axial Power T=k=1==ca, Z t

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OPERATIONA1. POWER IPIBAUUlt

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ENVELOPE FOR OPERATION FRC

(

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O'TO 115 (+ 10) EFPD

~

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UNIT 3

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3.5-23 l

"N NUCLEAR STATION I

Figure 3.5.2-3C1 Amendments Nos. 34, 34 & 31

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l h.4mM-e y% 6 is$ 9M 4

-64.

-,,., Y

l Power, 2 of 2568 MWt Restric\\

i ted Region

)

-16.47,102(;

) 16.83.102

-15.72.91' >

-90 4 5 6.83,91 1

-16.94.8 y

q>17.00,85 80 70

-23.64,64;

- 60 Permissible operating 50 Region

- 40

- 30 20 10 e

8 I

e i

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Power Inhalance, Z

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OPERATIONAL POWER IMBAUWCE ENVELOPE FOR OPERATION FROM 115 (+) EFPD to 226 (_+ 10) EFPI 3.5-23a UNIT _3 OCONEE NUCLEAR STATION Figure 3.5.2-3C2

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k.

Amendments Nos. 34, 34 & 31 i

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• N4 6-*

• • • N*-

b Power. I of 2568 MWt 1

Restricted Region

-25.5.102 -

4 6.71.102

-27.86,91 i) k

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)23.%.91

-80

-70

-60 Permissible Operating

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nesion

-50

-40

. -30 20 10 t

i e

I i

e i

i

-50

-40

-30

-20

-10 0

10 20 30 40 50 Arial Power Imbalance. %

y OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 226 (~+ 10) EFPD

~ UNIT 3 O

3.5-23b MN Figure 3.5.2-3C3 heendments Nos.

4. 34 & 31 O

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5 Lhit 1. Batch 4 g

12 i

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10 t

t i

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4 6

8 10 12 i

Axial Location of Peak Power From Bottom of Core, ft

3. %24 LUCA LDf1TED PRIDim ALLOWRBLE LINEAR HEAT RATE UETl

, MfwFF NUCLEAR STATION

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Figure 3.5.2-4

'ha

~

, heruksents Nos. 34. 34 1 31

1 Table 4.1-2

, MINIMUM EQUIPMENT TEST FREQUENCY

.lESP.

Test 7requency I

\\

Movement of Each Rod Bi-Weekly 1.

Control Rod Movement 2.

Pressurizer Safety Valves Setpoint 50% Annually 3.

Main Steam Safety Valves Setpoint 25% Annually 4.. Refueling System Interlocks Funce4an=1 Prior to Refueling II) 5.

Main Steam Stop valves Movement of Each Stop Monthly Valve Reactor Coolant System ( }

Evaluate Daily 6.

Leakage 7.

Condenser Cooling Water Functional Annually System Gravity Flow Test 8.

High Pressure Service Functional Monthly Water Pumps and Power

]'

Supplies 9.

Spent Fuel Cooling System Functional Prior to Refueling 10.

Hydraulic Snubbers on Visual Inspection Annually k

Safety-Related Systems High Pressure and Low ( }

Vent Pump Casings Monthly and Prior 11.

Pressure Injection System to Testing

12. Reactor Coolant System Flow Valid,-Oe Flow to be Once Per Yuel at least:

Cycle Unit 1 141.30 x 10 lb/hr Unit 2 141.30 x 10 lb/hr Unit 3 141.30 x 10 lb/hr l

(1) Applicable only when the reactor is critical o._

(2}- ap u,-M - only ubes the r==e*=e raa1==r is above 200*F and at a steady-state temperature and pressure.

(3) ap--=*4== ptsups==eindad-

)

I 4.1-9 Amendments Nos. 34, 34 & 31 w

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