ML19312C497

Amends 52,52 & 49 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Changing Tech Specs Re Limits for Unit 3,Cycle 3 Operation

ML19312C497
Person / Time
Site:	Oconee
Issue date:	11/21/1977
From:	Schwencer A Office of Nuclear Reactor Regulation
To:
Shared Package
ML19312C496	List: ML19312C495 ML19312C497 ML19312C500
References
NUDOCS 7912160059
Download: ML19312C497 (34)

Text

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UNITED STATES o

! ' 3, j' NUCLEAR REGULATORY COMMISSION F-E WASHINGTCN, D. C. 20555 E

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DUKE POWER COMPANY DOCXET NO. 50-269 OCONEE N'JCLEAR STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 52 License No. DPR-38 1.

The Nuclear Regulatory Commission (the Commission) has found tnat:

A.

The application for amencuent by Duke Power Company (the licensee) dated September 6,1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissior:'s rules and regulations set forth in 10 CFR Chapter I; B.

The 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 (i) that the activities authori;ed by this amendment can be conducted without endangering the health and safety of the puolic, and (ii) that such activities will be conducted in cocpliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical u the common defense and security or to the health and saft:y of tne public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements nave been satisfied.

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T 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.3 of Facility License No. DPR-38 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 52, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION RiW0L-

' A. 'Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: November 21, 1977

S UNITED STATES o

f p,7 ~ ' 7, NUCLEAR REGULATORY COMMisslON y'

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WASHINGTON, D. C. 20555

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%..... y DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 M4EiCMENT TO FACILITY OPERATING LICENSE Amendment No. 52 License No. OPR-47 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The apolication for amendment by Duke Power Company (the licensee) dated Septencer 6,1977, ccmplies with the standards and requirements of tne Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in ccnformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can ba conducted without endancering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amencment will not ce inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. OPR-47 is hereby amended to read as follows:

N (2) Tecnnical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. S2, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

~ JR THE NUCLEAR REGULATCRY CCP}i!SSION l

,, ua aae A. Schwencer, Chief Operating Reactors Branch fl Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: November 21, 1977 l

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UNITED STATES g

fo, NUCLEAR R5GULATORY COMMISSION

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.g W ASHINGTON, D. C. 20555 E

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DUKE POWER COMPANY OCCKET NO. 50-287 OC0 HEE NUCLEAR STATION, UNIT NO. 3 AMENDNENT TO FACILITY OPERATING LICENSE Amencment No. 49 License No. DPR-55, 1.

The Nuclear Regulatory Comission (the Commission) has found that:

The application for amendment by Duke Power Comoany (the licensee)

A.

dated September 6,1977, complies with the standards and requirements of the Atonic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The f acility 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 (i) that the activities authorized by this amendment can be conducted witnout encangering the health and safety of the public, and (ii) that such activities will be concucted in compliance with the Commission's regulations; D.

The issuance of tnis amenament will not De inimical to the conmon defense and sccurity or to the health and safety of the puDliC; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

v-3 s

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility License No. OPR-55 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 49, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY CCMMISSION

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A. Schwencer, Chief Operating Reactors Branch 41 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: November 21, 1977 O

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J ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 52 TO DPR-38 AMENDMENT NO. 52 TO DPR-47 AMENDMENT NO. 49 TO DPR-55 DOCKET N05. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages.

2.1-3c 3.5-10 2.1-3d 3.5-11 2.1-6 3.5-16 2.1-9 3.5-16a 2.1-12 3.5-17 2.3-1 3.5-20 2.3-3 3.5-20a 2.3-4 3.5-20b 2.3-7 3.5-23 2.3-10 3.5-23a 2.3-13 3.5-23b 3.5-9 4.1-9 2.

Add pages 3.5-231, 3.5-23j and 3.5-23k

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Bases - Unit 3 The safety 1Laits presented for Oconee Unit 3 have been generated using 3AW-2 critical heat flux correlation (l) and the Reactor Coolant System flow rate of 6 lbs/hr for four-pump operation).

l 106.5 percent of the design flow (131.32 x 10 The flow race utilized is conservative compared to the actual measured flow rate.(2)

To maintain the integrity of the fuel cladding and to prevent fission produce 1

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 coef ficient is i

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 j

Although DN3 is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and 7ressure can be relat,ed to DN3 through the use of the BAW-2 correlation (l).

The 3AW-2 correlation has been developed to predict DN3 and the location of Dh3 for axially uniform and non-uniform heat flux distributions.

The local DNB ratio (DNER), defined as the ratio of the heat flux that would cause DN3 at a particular core location to the actual heat flux, is indicative of the margin to DN3. The mini =um value of the DN3R, 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 DN3 will not occur; this is considered a conservative sargin 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 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1C represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible ther=al power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 139.86x 106 lbs/hr.).

This curve is based on the following l

nuclear power peaking factors with potential fuel densification and fuel rod bowing effects:

F

= 2.67; F

= 1.78; 7

= 1.50.

The d-2 sign peaking q

AH z

ceabination results in a more conservative DNBR than any other power shape that exists during normal operation.

The curves of Figure 2.1-2C are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod

bowing, j

N = 2.67 or 1.

The 1.30 DNBR limit produced by a nuclear peaking factor of Fg the combination of the radial peak, axial penk and position oc the axial peak that yields no less than a 1.30 DN3R.

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2.1-3c i

1 l

Amendments Nos. 52, 52 & 49 i

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2.

The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.15 kw/ft for Unit 3.

Power peaking is not a directly observable quantity, and, therefore, limits have been established on the bases of the reactor power imbalance produced by the power peaking.

The specified flow rates for Curves 1, 2 and 3 of Figure 2.1-2C correspond to the expected minimum flow rates with four pumps, three pumps and one pump j

in each loop, respectively.

The maximwn thermal power for three-pump operation is 85.3 percent due to a power level trip produced by the flux-flow ratio 74.7 pe_rcent flow x 1.055=

78.8 percent power plus the mavisum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.

For each curve of Figure 2.1-3C a pressure-temperature point above and to the left of the curve would result in a DN3R greater than 1.30 or a local quality at the point of minimum DN3R less than 22 percent for that particular reactor coolant pump situation. The curve of Figure 2.1-lC is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3C.

References (1) Correlation of Critical Heat Flux in a Sundle Cooled by Pressurized Water, BAW-10000, March 1970.

(2) Oconee 3, Cycle 3 - Reload Report - 3A'4-1453, August, 1977.

l l

l l

l 2.1-3d Amendments Nos. 52, 52 & 49

f g

2400 2200 ACCEPTABLE OPERATION a

5 3

2000 a.

=

5 UNACCEPTABLE OPERATION e

3 1800 1600 560 580 600 620 640 Reactor Coolant Outlet Temperature,F CORE PROTECTION SAFETY LIMITS UNIT 3 k OCONEE NUCLEAR STATION

%1 2.1-6 Figure 2.1-lC I

i Amendment Nos. 52, 52 & 49

THEasAL PosEA LEVEL, 1

. 120

-37,112 37,112

- 110

-- 100 ACCEPTABLE 4-PUMP OPERATION 90

-37,35.3 37,85.3

-. 80

-52,80 49.2,80 ACCEPTABLE 3 & 4

-- 70 PUMP OPERATION

-37,58.2

__ 60 37,58.2 52,53.3 50 40 ACCEPTABLE 2,3 & 4

- 30 PUMP OPERATION

-52,26.2 49.2.26.2 g

10 l

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I 1

60

-40

-20 0

20 40 60 Reactor Power imoalance, 5 Curve Reactor Coolant Flow, gpm 1

374,880 (100%)*

2 280,035 (74.77.)

3 183,690 (49.0%)

CORE PROTECTION SAFETY LIMITS UNIT 3

• 106.5% of ftrst-core design flow.

b 2.1-9 l sui n** s; OCONEE NUCLEAR STATION W

Figure 2.1-2C 4endment Nos. 52, 52 & 49

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2400

/

/

ACCEPTABLE OPERATION 2200 a.

5 E

2000 7

0 1

2 3

5 f

/

O" 1800 1600 1

560 580 600 620 640 Reactor Coolant Outlet Temperature,F

Power, Pumps Type of Curve Coolant Flow, gpm i

Operating Limit 1

374,880 (100%)*

112 4

DNBR 2

280,035 (74.7%)

86.7 3

DNBR 3

183,690 (49.0*)

59.0 2

Quality

• 106.5% of first-core design flow.

CCRE PROTECTI0ti SAFETY LIMITS UtlIT 3

1-l' 4

l I

OCONEE NUCLEAR STATION rigure 2.1-3C Amendment Nos. 52, 52 & 49 l

W 2.3 LIMITING SAFETY SYST22d SETTINGS, PROTIC*I7E INSTRL* MENTATION Applicability Applies to instruments monitoring reactor power, reactor power i= balance, reactor coolant system pressure, reactor coolan: outlet :emperature, flow, number of pumps in operation, and high reactor building p: essure.

Obiective To provide automatic protective ac: ion to prevent any combina: ion of process variables frem exceeding a saf ety 1121:.

5:ecification The reactor protec:ive system crip set:ing li=its and :he permissible 1

bypasses for the instrument channels shall be as s:a:ed in Table 2.3.lA-Uni:

and 2.3-13-Uni: 2 2.3-lC-Unit 3 Figure 2.3-2A-Uni: 1 2.3-23-Uni: 2 2.3-2C-Uni: 3 The pump =cni: ors shall produce a reac:or crip for :he following condi:icns:

Loss of one pu=p during four-pump operatien if power level is greater a.

than 30% of rated pcwer Lose of two pumps and reactor power level is greater than 55% of ra:ed b.

pcwer. (Power /RC pump crip se: point is reset to.55% for operation wi:h one pump in each loop).

Loss of two pu=ps La one reac:ce coolant loop and reac:or power level is c.

3: eater :han 0.0% of raced. power.

d.

Loss of one or two pumps during two-pu=p operatica.

Bases The reac:or protec:ive systes consis:s of four Lastrumen: channels :o =eni:or each of several selec:ed plant condi:1ons which will cause a reactor trip if any one of these conditions devia:es frem a pre-selec:ed opera:ing range :o the degree : hat a safety lisi: say be reached.

The trip setting limi:s for protec:1ve sys:em instru=enta: ion are listed in Table 2.3-LA-Uni: 1.

The saf ety analysis has been based upon -hese pro:ec:1ve 2.3-13-Uni: 2 2.3-lC-Uni: 3 sys:em instrumentation : rip setpoints plus calibra:ica and ins:ru=enca:1on errors.

Nuclear Overcover i'

A reac:or trip at high pcwer level (neucron flux) is provided :o preven:

da= age to the fuel cladding from reac:ivity excursiens co rapid cc be detected by pressure and :empera:ure seasuremen:s.

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2.3-1 Acendgent Nos. 52, 52 & Jo 1

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level trip and associated reactor power / reactor power-imbalance boundaries by 1.0 551 for a 1 flow reduccion.

The pcwer-to-ficw reduction ratio is 0.949 during single loop operation.

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Pumo Monitors The pump monitors prevent the minimum core CNBR from decreasing below 1.3 by tripping the reactor due :o the loss of reac:or coolant pump (s).

The circuitry monitoring pump operational sca:us provides 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 restric: the power level for :he number of pumps in operation.

Reactor Coolant Svs:em Pressure During a startup accident from low power or a slow red wi:hdrawal from high power, the system high pressure set poin is reached before the nuclear over-power trip se point.

The trip set:ing limi: shown in Figure 2.3-1A - Uni: 1 2.3 Uni: 2 2.3-IC - Unit 3 for high reac:or coolan: system pressure (2355 psig) has been established to

=aintain the system pressure below :he safe:y limi: (2750 psig) for any design transient. (1)

I The low pressure (1800) psis and variable low pressure (11.14 T

-4706) trip out (1300) psig (11.14 Tout-4706)

(1800) psig (11.r4 Tout-4706) setpoints shown in Figure 2.3-1A have been established to maintain the CNB 2.3-13 2.3-LC ratio greater than or equal :o 1.3 for those design accidents that resul: in a pressure redue:1on. (2,3)

Due to the calibration and instrumenta:1on errors the safe:y analysis used a l

variable low reac:or coolant system pressure : rip value of (11.14 T

-4746) out (11.14 Tou: -4740) l (11.14 Tou: -4746) l Ccolant Cu:let Temeersture a.c high reactor coolant outle: temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-13

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2.3-lc temperatures in the operating range.

Que to calibracion,and instrumentation errors, the safety analysis used a trip set point of 620 F.

Reactor Building Pressure The hign reactor building pressure trip set:ing limi: (4 psig) provides positive assurance that a reactor trip will occur in :he unlikely event of a loss-of-coolant accident, even in the absence of a icw reactor coolant system pressure trip.

2.3-3 Amendment Nos. 52, 52 & 49

Shutdown 3veass In order to provide for control rod drive tests, tero power physics testing.

and startup procedures, there is provision for bypassing certain segments of The reactor protection system segments which the reactor procaction system.

can be bypassen are shown in Table 2.3-LA.

Two conditions are i= posed when 2.3-13 2.3-1C the bypass is used:

point must be By administrative control the nuclear overpower trip set 1.

reduced to a value 1 5.0% of rated power during reactor shutdown.

2.

A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.

The purpose of the 1720 psig high pressure trip set point is to prevent normal This high operation with part of the reactor protection system bypassed.

pressure trip set point is lower than the normal low pressure trip set point The over be tripped before the bypass is initiated.

so that the reactor must power trip set point of 1 5.0% prevents any significant reactor power from Sufficient natural being produced when performing the physics tests.

circulation (5) would be available to remove 5.0% of rated power if none of the reactor coolant pumps were operating.

Two pump Operation A.

Two Loop Operation Operation with one pump in each loop will be allowed c.nly f ollowing.

reactor shutdown. After shutdown has occurred, reset the pump contact monitor power level trip setpoint to 35.00.

D*PD 7D ^Tl &

is g@

d

. $.l1 is 3.

Single Loop Operation Single loop operation is permitted only after the reactor has been Af ter the pu=p contact =enitor trip has occurred, the following tripped.

actions Jill permit single loop operation:

to 55.0.

the pump contact eenitor power level trip setpoint 1.

Reset Trip one of the two protective channels receiving outlet te=perature 2.

information from sensors it. the Idle Loop.

3.

Reset flux-flow setpoint to 0.949.

l RE7ERENCES (1) FSAR, Section 14.1.2.2 (4) FSAA, Section 14.1.2.3 (2) FSAR, Section 14.1.2.7 (5) FSAa, Section 14.1.2.6 (3) FSAR, Section 14.1. 2.8 l

2.3-4 i

Amendment Nos. 52, 52 & 49

2400 T = 619'F P = 2355 psig 2300 T

2200 E.

ACCEPTABLE

.i OPERATION E

C" 2100 Z2 O

a UNACCEPTABLE 8

a OPERATION

-f 2000 E

s' s-s 1900 -

a.

P = 1800 psig 1800 T = 584F I

I I

I 540 560 580 600 620 S40 Reactor Outlet Temperature, F

PROTECTIVE SYSTEM MAXIML'M ALLOWABLE SETPOINTS i

l 2.3-7 UNIT 3 b\\

bt nwn; OCONEE NUCLEAR STATION Y

l Figure 2.3-lC Amendment Nos. 52, 52 & 49

THERMAL POWER LEVEL 5

- 120 UNACCEPTABLE OPERATION

-26.1,105.5 25.0,105.5

+

ACCEPTABLE 100 s

l4 PUMP l

g p

N OPERATION e

90 e,

l26.1,78.8 25.0, 78.8 80

-40.1,73.5 ACCEPTABLE 36.9,73.5 70 3 & 4 PUMP OPERATION 60 I

I

-26.1.51.7 25.0.51.7 i

-40.1,46.8 ACCEPTABLE 2,3, & 4 i

40 i

PUMP OPERATION 30 36.9,19.7

-40.1,19.7 20 i

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l 10

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i 60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 l

Power Isaalance, 5 PROTECTIVE SYSTEM MAXIMUM ALLONABLE SETFCINTS UNIT 3 2.3-10

, OCONEE NUCLEAR STATION Figure 2.3-2C Amendment Nos. 52, 52 & 49

Table 2.3-IC Unit 3 Reactor Protective System Trip Setting Limit _s One Reactor Four Reactor Three Reactor Coolant Pump Coolant Pumps Coolant Pumps Operating in Operating Operating Each Loop (Operating Power (Operating Power (Operating Shutdosa RPS Segment

-100% Ra ted)

-75% Rated)

-49% Rated)_

Bypass 1.

Nuclear Power nix.

105.5 105.5 105.5 5.0

't

(% Rated) 2.

Nuclear Power Max. Based 1.055 times flow 1.055 times flow 1.055 times flow Bypassed l

on Flow (2) and Imbalance, minus reduction minus reduction minus reduction

(% Rated) due to imbalance due to imbalance due to imbalance F

3.

Nuclear Power flax. liased NA 80%

55%

Bypassed I

on Pump Monitors, (% Rated)

I4) 4.

High Reactor Coolant 2355 2355 2355 1720 Syste:S Pressure, psig, Max.

5.

l.ow Heactor Coolant 1800 1800 1800 Bypassed System Pressure, psig Min.

6.

Variable Low Reactor (11.14 T

-4706)

(11.14 T - 4706)

(11.14 T

-4706)

Bypassed l

u out out out g

Coolant S,ystem Pressure ct psig, Min.

$5 7.

Reactor Coolant Temp.

619 619 619 619

.g F., Max.

m 8.

liigh Reactor Building 4

4 4

4

<n

,m Pressure, psig, Max.

l0 a.

(1) T is in degrees Fahrenheit ( F).

,g (2) Reactor Coolant System Flow, %.

(3) Administratively controlled reduction set only during reactor shutdown.

(4) Automatically set when other segments of the RPS are bypassed.

?

pump cp3rctius. Also, excupting physics test:

e exsrcising c:ntrol rods, tha cxial powar shaping control rod insortion/

withdrtwal limits are spscified on figuras 3.5.2-4A1. 3.5.2-4A2 and 3.5.2-4A3 (Unit 1), 3.5.2-431, 3.5.2-432, and 3.5.2-433 i

j (Unit 2), and 3.5.2-4C1, 3.5.2-4C2, and 3.5.2-4C3 (Unic 3).

l

_r If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control red position. An acceptable control rod posi:1on shall then'be attained within two hours. The minimum shutdevn margin-j required by Specification 3.5.2.1 shall be =aintained at all times.

d.

Except for physics tests, power shall not be incre'ased above the power level cutoff as shown on Figures 3.5.2-1A1, 3.5.2-1A2 (Uni: 1), 3.5.2-131, 3.5.2-132, and 3.5.2-133 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, 3.5.2-1C3 (Unit 3), unless the following requirements are met.

(1) The xenon reactivity shall be within 10 percent of the value for operation at steady-sta:e rated power.

(2) The renon reactivity.wot:h has passed i:s final =ax1=t:n or mini =::s peak during its approach to 1:s equilibrium value for operaticit at the pcuer level cutof f.

3.5.2.6 Reactor power i= balance shall be =enitored on a frequency not to

. exceed two hours during power operation above 40 percent ra:ed power.

Except for physics tests, imbalance shall be maintained wi:hin the envelope defined by Tigures 3. 5.2-3A1, 3. 5. 2-3A2, 1. 5. 2-3A3, 3. 5. 2-331 l

3.5.2-332, 3.5.2-333, 3.5.2-3C1, 3.5.2-3C2, and 3.5.2-3C3.

If the i=-

balance is not within the envelope defined by these figures, corrective measures shall be taken to achieve an acceptable imbalance.

If an accep-table imbalance is not achieved within two hours, reactor power shall be reduced until i=balanc'e limits are =et.

3.5.2.7 The con:rol rod drive patch panels shall be locked a: all ci=es wi:h lisi:ed access :o be authorized by the manager or his designated al:ernate.

3.5.2.8 For oconce Unit 1, in the event Specifica:icns 3.5.2.4.a ei 3.5.2.5.c-are not =ct, operation shall be restricted as fo11cws:

The core thermal powcr shall bc li=ited to 75 percent full power.

a.

b.

The nuclear power naximu= setpoint shall be S4 percent full.pcicr.

s c.

The quadrant til: shirl not exceed 6.03 percen:.

d.

The.r ceul atin2 centrol rod insertion /t:1.:' dra cal lini:s ar'c specified an Figure 3.5. 2-6Al-If any of the abcVe provisions are nc.c =ct trichin tiro hours, the ren::er shall be in :.he hot shutdotin condi,:icn wi:hin an addi:fonal 4 hcurs.

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Within 25 E7?D of the da:c of issuance of this Specification, previde a repot: and analysis of the quadrant' flux til: cbsc-ved and projec:icns for the next 25 Er?D. Operatien above 75% is not authori:cd if flux tilt is above 3.41% ' unless an amend = cat request is submitted accompanied l

by detailed evaluation and justifientica 3.5-9 i

Amendment Nos. 52, 52 & ao f

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Bases The power-imbalance envelope defined in Figures 3. 5.2-3A1, 3. 5.2-3A2, 3. 5.2-3A3, 3.5.2-381, 3.5.2-332, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2 and 3.5.2-3C3 is based on l

LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-5) 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 would 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 power spike f actors (Units 1 and 2 only)

I d.

Hot rod manuf acturing tolerance factors e.

Fuel rod bowing power spike f actors l

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

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

Group Function 1

Safety 2

Safety 3

Safety 4

Safety 5

Regulating 6

Regulating 7

Xenon transient override 8

APSR (axial power shaping bank)

The rod position limits are based on the most 11=iting 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 mini =um available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any l

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 0.65%

will not coatain single rod worths greater than ak/k at rated power.

These values have been shown to be safe a

by the safety analysis (2, 3, 4, 5) of the hypothetical rod ejection accident.

A maximum single inserted control rod worth of 1.0% ak/k is allowed by the rod position limits at hot zero power.

A single inserted control rod worth of 1.0% ak/k at beginning-of-life, hot zero power vould result in a lower transient peak thermal power and, therefore, less severe environmental consequences thanrated powe 0.65% ak/k ejected rod worth at a

l or excore detectors

• • Actual operating limits depend on whether or not incore are used and their respective instrument calibration errors. The method used to define the operating Itaits is defined in plant operating procedures.

- 3.5-10 APendment Ngs. 52, 52 & 4o

.m 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 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 5.10% for Unit 3 are measurement system independent. The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each system are defined in the station operating procedures.

measurement The quadrant tilt and axial imbalance sonitoring in Specification 3.5.2.4 and 3.5.2.6, respectively, normally v111 be perf ormed in the process computer.

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

Allevance 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 i= balance sust 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 to prevent excessive po,ver peaking by transient xenon. The xenon reactivity cust,be beyond its rinal =aximum or mini =um peak and approaching its equili-brium value at the power level cutoff.

REFERENCES 175AR, Section 3.2.2.1.2 2ISAR, Section 14.2.2.2

' FSAR, SUPPLEMENI 9 I

4B&W RTEL DENSITICATION REPORT 1AW-1409 (UNIT 1) 1AW-1396 (UNIT 2)

EMW-1400 (UNIT 3) 5GCONEE UNIT 1, CYCLE a REI.0AD REPORT, l

EAW-1447, March, 1977 Section 7.11 l

Amendment Nos. 52, 52 & 49

d t13.102 174, t,302 a o232.0,102 RESTRICTED REGION 100 -

OPERATION IN THl3 30 REGION 13 NOT 174.1.90 232.0,90 POWER LEVEL ALLOWED OUTOFF 161.2,30 251.4,20 80 SHUTOOWN 150.70 EARGIN 300,70 70 Lisit g

N 80.60 36,50

• , 50 5

'8 PERILS $1BLE OPIRATING REGION 30 20

,15 10 0,0 t

I f

f i

a f

1 l

I 1

i 0

20 40 60 80 100 120 140 150 180 200 220 240 260 280 300 Roa Ina 1, 5 litnsraen 0

25 50 75 100 0

25 50 75 100 i

e i

i i

e e

1 l

Graap 7 Group 5 0

25 50 75 100 i

i i

r Group 6 R00 POSITION LIMITS FOR FOUR PUMP OPERATION FRCM 0 TO 100 (+ 10) EFPD UNIT 3 b\\

btnam OCONEE NUCLEAR STATION 3.5-16

.g Ficure 3.5.2-1C1 i

Amendment Nos. 52, 52 & 49

174.1,102 o 232.0.102 100 RESTRICTED

~

REGION i l74.1.90 32.0,90 90 OPERATION IN THl$

pg,(g g(,

REGION l$ NOT ALLOWED CUTOFF 80 161.2.80 251 4,80 i

70 300,70

$ NUT 009N MARGlN LlEIT 60 50 108,50 2

~

PERMIS$1SLE OPERATING 30

~

REGION 20 70,15 80,15 0,8. 2 to O'!

0 20 40 60 80 100 120 140 160 180 200 220 240 250 290 300 Roa Index, 5 litnaraan 0

25 50 75 100 0

25 50 75 100 i

Group 5 Group 7 0

25 50 75 100 i

i Group 6 R00 POSITION LIMITS FOR FOUR PUMP CPERATION FRCM 100 (i 10)

EFPD TO 235 (t 10) EFP0 UNIT 3

',e OCONEE NUCLEAR STATION 3.5-16a Y

Figure 3.5.2-lC2 l

l Amen &ent Nos. 52, 52 & 49 l

251.4,102 208.102 o

y,g _

N 251.4.90

,e 90 -

OPERATION IN THl3 REGION l$ N07 ALLOWE0 241.7.80 to -

~

SHUT 00fM MARGIN LIMIT 3

60

/

o 0

IU.M PERMISSIBLE OPERATING REGION 3g 40 30 20 190.15 110.15 0.5.9 10 0.0 0

20 40 80 80 100 120 140 160 180 200 220 240 250 280 300 Roo inces. 5 fitnarasn 0

25 50 75 100 0

25 50 75 100 i

e i

i Group 7 Group 5 0

25 50 75 I00 Group 6 R00 POSITION LIMITS FOR FOUR FUMP OPERATION AFTER 235

(+ 10) EFPD UNIT 3 loatMet9 OCONEE NUCLEAR STATION I

3.5-t7 Figure 2.5.2-1C3 Amendment Nos. 52, 52 & 49 l

l

.)

113.102 161.2.102 251.4.102 100

/

RESTRICTED OPERATION IN THIS 90 g@

t 150,89 P' s 300.89 REGION 15 NOT ggk OPERATION ALLORED 3

0,75 E 70 j

SHUTDOWN

$ 60

- MARGIN j

Lluli W 50 38,50 PERutSSIBLE OPERATING REGION 40

=

3-- 30 E,0

,s 15 j 10 0.0 i

1 0

20 40 60

• 10 120 140 160 180 200 220 240 260 280 300 Roa inces 5 fitndrawn 0

25 50 75 100 0

25 50 75 10E i

Group 5 Group 7 0

25 50 75 100 1

Group 6 i

ROD POSITION LIMITS FOR TWO-AND THREE-PUMP OPEPATION FROM 0 TO 100 (I 10) EFPD UNIT 3 h' M gW'; OCONEE NUCL 3*5-20 at n=t e Figure 3.., C1

.c-c Amendment Nos. 52, 52 & 49

174 1.10 251.4,102 IE ON OPERAtl0N IN THl3 REGION p p 90 5

15 NOT ALL0 SED 300.89 2 80

$ 70 E

SHUTOOWN MARGIN LIMIT PERMIS$1BLE OPERAflhG REGION 80 3

}50 108.50 t

2: 40 3

[ 30 E 20 2

70,15 0,8.2 10 ESTRIOTED FOR 2 & 3 PUuP OPERATION

, 0. 0 0

20 40 60 80 100 120 140 160 ISO 200 220 240 260 280 300 Roa inces, 5 titnarson 0

25 50 15 100 0

25 50 75 100 i

i i

Group 5 Group 7 0

25 50 75 100 i

Group 6 R00 POSITION LIMIS FOR TWO-AND THREE-PUMP OPERATION FROM 100 (

10) TO 235 (; 10) EFP0 UNIT 3 OCONEE NUCLEAR ;TATION 3.5-2Ca w

Figure 3.5.2-2C2 Amendment Nos. 52, 52 & 49 l

-s RESTRICTED FOR 3 PUMP QPERATION 100 OPERATION IN TH15 REGION is NOT ALL0sEO 90 196,92 60 h70 SHUT 00fN MARGIN LIMIT i

u 60 a

3 W to 142.50 PERMISStatt OPERATING REGION to.

N

=

e a:*

30

=

e-l 20 100.;5 I 10 0.5.9 2

STRICTED FOR 2 & 3 PUMP OPERAfl0N t

0 20 40 60 60 100 120 140 ISO 180 200 220 240 260 280 300 Rod inces. 5 litnoraen 0

25 50 75 100 0

25 50 75 100 l

t e

e f

1 e

Group 5 Group 7 0

25 50 75 100 Group 6 1

1 R00 POSITION LIMITS FOR TWO-AND THREE-PUMP OPERATION 1

AFTER 235

(+_ 10) EFPD 1

UNIT 3 b

l 3.5-20b

'pii nm; OCONEE NUCLEAR STATION W

Figure 3.5.2-2C3 Artendar.t Nos. 52, 52 & 49

(

Power, 5 of 2568 n t RESTRICTED REGION

^ 8.66,102

-13.73,102

, i gg

^

-23.54,90 90 11.35,90 80

)12.63,80 26.66,80 <

70 60 50 40 PERMISSIBLE OPERATING REGION 20 10 i

i i

I 1

e t

50 40 30

-20 10 0

10 20 30 40 50 Axial Power imoalanca, 5 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FRCM 0 TO 100 (+ 10) EFPD UNIT 3 b

3.5-23 Sum's OCONEE NUCLEAR STATION W

Figure 3.5.2-3Cl Amendment Nos. 52, 52 & 49

Power, 5 of 2568 n t RESTRICTED REGION

^

25.57,102 Q

-100 i

90

) 11.35,90

-24.65,90:D 26.11,80 (

80

<> 12.63,80 70

=

60 PERWISSIBLE 50 i

OPERATING REGION 40 30 20 10 e

i i

i i

i i

i t -w 50 40 30

-20

-10 0

10 20 30 40 50 Axial Power imostance, 5 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATI0ti FROM 100 (+) 10 EFFD TO 235 _(;,10)

EFPD t.tJIT 3 OCONEE NUCLEAR STATION

(

Figure 3.5.2-3C2 Amendment Nos. 52, 52 & 49 l

l

Power, 5 of 2568 M t RESTRICTED REGION

-27.19,102 'r

-100 T

. 90 o 18.42,90

-29.94,90 ( )

30 c) 10.38,80

-29.94,80 0

. 70 PERillSSIBLE 60 OPERAT ING REGION

. 50

. 40 30 20 10 t

i e

i i

e i

i I

i 50 40

-30 20

-10 0

10 20 30 40 50 Axial Power imoalance, 5 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 235 (+ 10) EFPD UNIT 3 e

OCONEE NUCLEAR STATION 3.5-23b Figure 3.5.2-3C3 Amend..ent Nos. 52. E2 & 49 i

6.6.102 33.3.102 RESTRICTED 100 REGION 90 0,90 38.4,90 j

RESTRICTED 64.4,80 80 REGION 70 100,70 60 iG g

50

==

PERNISSIBLE m

40 g

OPERATING REGION 20 10 0

~

0 10 20 30 40 50 60 70 80 90 100 APSR, 5 litnerawn APSR POSITION LIMITS FOR OPERATION FROM 0 TO 100 1 10 EFPD 3.5-231 :

OCON E NUCLEAR STATION Figure 3.5.2-4Cl Amendment Nos. 52, 52 & 49

Figure 8-16.

APSR Position Limits for Operation From 100 :: 10 to 235 2 10 EFPD - Oconee 3, Cycle 3 100 RESTRICTED REGION 51.4,90 90 64.4,80 80 70 100,70 G"

60 3

N 50 e

d 40 a

n.

PERillSSIBL E 30 OPERATING REGION 20 10 0

O 10 20 30 40 50 60 70 80 90 100 APSR, 5 fitnaraun APSR POSITION LIMITS FOR OPERATION FROM 100 t 10 TO 235 1 10 EFPD UNIT 3 3.5-23j

(

OCONEE NUCLEAR STATION 3.5.2-4C2 Amendment Nos. 52, 52 & 49

~

s Figure 8-17.

APSR Position Limits for Operation Af ter 235 2 10 EFPD - Oconee 3, Cycle 3 36.3.102 RESTRICTED 100 REGION 51.4.90 90 64.4,80 80 70 100,70 E

60

=N PERMISSIBLE OPERATING REGION 50 f.

40 2

30 20 10 0

i i

i i

i 0

10 20 30 40 50 60 70 80 90 100 APSR, 5 fathdrawn APSR POSITION LIMITS FOR OPERATION AFTER 235 + 10 EFPD UNIT 3 3.5-23k 3

) OCONEE NUCLEAR STATION Figure 3.5.2 l.C3 Amendment Nos. 52, 52 & 49

]

.s j

Table 4.1-2 MINIMtM ICU!? E 7 TIST ??.ECUI' ICY Item Test Frecuenev Concrol Rod Movenen (1)

Move =en: of Each Rod 31-Weekly 1.

2.

Pressuri:er Safety Valves Se poin:

50% Annually 3.

Main Stea= Safety Valves Se: point 25% Annually 4.

Refueling Systes interlocks Fune:1cnal Prior ec Refuella; 5.

Main S:es= Stop Valves Movement of Each stop Mon:hly II)

Valve

()

6.

Reactor Coolan: Systa=

Evaluate Daily Leakage 7.

Condenser Cooling Wa:er Functienal Annually Systa= Gravi:y Flow Tes 8.

High Pressure Service Functional Monthly Wa:er Pu=ps and Power Supplies e

Prior to 9

Spen: Fuel Cooling Systa=

Functional Refueling l

igh Pressure and Lov(3)

Ven: Pu=p Casings Monthly and ?;1c;

'to Testing.

10* Pressure Injec:1cs Systen i

f j

.(11 Applicable only when e.he reactor is critical o

~

(2)

Applicable only when the reac:c; coolann is above 200 F and a: a steady-s:a:e :e:pera:ure and pressure.

l

~

(3) Opera:ing pu=ps excluded.

Amendment Nos. 52, 52 & 49

. 1-9 D

• lD lD

' T 313 mom enlh

\\