ML19312B890

Proposed Tech Specs 2.1,2.3 & 3.5 Re Reactor Core Safety Limits,Protective Instrumentation Limiting Safety Sys Settings & Control Rod Group & Power Distribution Limit

ML19312B890
Person / Time
Site:	Oconee
Issue date:	03/30/1977
From:	DUKE POWER CO.
To:
Shared Package
ML19312B884	List: ML19312B883 ML19312B890
References
NUDOCS 7911250079
Download: ML19312B890 (28)

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2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.

Objective To maintain *:he integrity of the fuel cladding.

Specification The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1A-Unit 1. If the actual pressure / temperature point is below 2.1-1B-Unit 2 2.1-lC-Unit 3 and to the right of the line, the safety limit is exceeded.

The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2A-Unit 1. If the actual reactor-thermal-power / power 2.1-2B-Unit 2 2.1-2C-Unit 3 imbalance point is above the line for the specified flow, the safety limit is exceeded.

Bases - Unit 1 The safety limits presented for Oconee Unit 1 have been generated using BAW-2 critical heat flux (CHF) correlation (1) . The reactor coolant system flow rate l utilized is 106.5 percent of the design flow (131.32 x 10'f lbs/hr) based on four-pump operation.(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 i regime is termed " departure from nucleate boiling" (DNB). At this point, l there is a sharp reduction of the heat transfer coefficient, which would result I in high cladding temperatures and the possibility of cladding failure. Although l DNB is not an observable parameter during reactor operation, the observable '

parameters of neutron power, reactor coolant flow, temperature, and pressure Q;O

can be related to DN3 through the use of the BAW-2 correlation (1). The 3AW-2 correlation has been developed to predict DNB and the location of DN3 for axially uniform and non-uniform heat flux distributions. The local DN3 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 margin to DNB. The minimum 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 confidence level that DNB will not occur; this is considered a conservative =argin 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 itsits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setponts to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1A represents the conditicas at which a mini =um DNBR of 1.30 is predicted for the maxi =u= possible thermal power (112 percent) when four reactor coolant pumps are operating (mini =um reactor coolant flow is 106.5 percent of 131.3 x 100 lbs/hr.). This curve is based on the combination of nuclear power peaking factors, with potential ef fects of fuel l densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.

The curves of Figure 2.1-2A are based on the = ore restrictive of two thermal limits and include the effects of potential fuel densification and rod bowing:

1. The 1.30 DNBR limit produced by the combination of the radial peak, axial peak and position of the axial peak that yields no less than a 1.30 DN3R.

2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The li=it is 20.15 kw/ft for Unit 1.

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, 3 and 4 of Figure 2.1-2A correspond to the expected minimum flow rates with four pumps, three pumps, one pump in each loop and two pumps in one loop, respectively.

The curve of Figure 2.1-LA is the mosc restrictive of all possible reactor coolant pump-maxi =um thermal power ccabinations shown in Figure 2.1-3A.

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

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

l l

l 2.1-2

For Figure 2.1-3A, a pressure-tempsrature point above and to tha left of

, thy curva would result in a DNBR greater than 1.30.

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

i l-(2) Oconee 1, Cycle 4 - Reload Report - BAW-1447, March, 1977. l j

I 4

l i

j l

r 4

4 1

1 l'

k

.I J

2.1-3 4

Y E

- ----we , - - * , , - -e--- p ,-+-- r- --- v e r- , --,w - ~ 4 .

cre.-e-a-w,- a g gy - --v---s, p q r-y--ye.- r-=g- -m

~

2400

o. 2200 '

2 5

U E 2000

/

/

5 1800 r l

l I600 - - . .

560 580 600 620 640 Reactor Coolant Outlet Temperature,F CORE PROTECTION SAFETY 2.1-2

- LIMITS. Ut:IT 1

,b\

.avig non, OCONEE NUCLEAR STATION W r igure 2.1-1A

8 f

THERNAL POWER LEVEL, S

_ _ 120

(-17.2.112) '

-- 110 ACCEPTABLE 4

~~ 100 pg,, t

( 40.98)

OPERATION

__ 90

__ 80 (50.80)

ACCEPTABLE 2

( 40,72)

__ 70 3 & 4 PLW OPERATICN

__ 60 3

-- 50 ACCEPTABLE 4

(-40,44.2) 2,3 g 4

-- 40 Pulf OPERATION

_._ 30 (50.26.0)

__ 20

-- 10 1  ! f 1 1 I 40 20 0 +20 +40 +60 60 Reactor Paeer Imaalance, ,

CURVE REACTOR COOLANT FLOW (GPM) 1 374880 2 280035 3 1e3690 4 204310 l

CORE PROTECTION SAFETY LIMITS, UNIT 1 lontn OCONEE NUCLEAR STATION

.1-7 M Figure 2.1-2A

i

. , l l

2400 EPTABLE

. 2200 OPERAil0N f

? 7 2

5 I 2000 /

2 g 1 o

O 1800 m I r 1600 560 580 600 620 640 Reactor Coolant Outlet Temperature-F CURVE REACTOR COOLANT FLOW (GPM) POWER PUMPS OPERATING (TYPE OF LIMIT) 1 374880 (1005)* 1125 4 (DNBR) 2 280035 (74.75) 86.75 3 (DNBR) 3 183690 (49.05) 59.05 2 LOUALITY)

' 106 55 0F FIRST CORE DESIGN FLOW CORE PROTECTION SAFETY LIMITS, UNIT i

'o Al OCONEE NUCLEAR STATION 2 .-10 DI Figure 2.1-3A

. . . , ~

During nor=al plant oper.'eion with all t eactor coolant p 's operating, reactor trip is initiate .sh:n thn ratctor powar level reaches 105.5% of rated powar. . Adding to this ths possible variation in trip setpoints due to cal,1bration and instrumtnt errors, the maximum 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 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 acco==odate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has de=onstrated that the specified i power-to-flow ratio is adequate to prevent a DN3R of less than 1.3 should a low flow condition exist due to any electrical malfunction.

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 coolanc flow rate decreases. The power level trip set point produced by the power-to-flow ratio provides overpower DN3 pro-tection for all modes of pump operation. For every flow rate there is a =axi-sum 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 coolant pumps are operating if power is 105.5% and reactor flow rate is 100%, or flow rate is 94.8% and power level is 100%. l

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 l level is 75%.

3. Trip would occur when two reactor coolant pe=ps are operating in a single loop if power is 51.7% and the operating loop flow ra:e 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 31.7% and reactor flow rate is 49.0% or flow rate is 46.4% and the power level is 49%.

The flux-to-flow ratios account for the maximum calibration and instrument 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.

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

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The reactor power imbalance (pewer in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio such that the boundaries of Figure 2.3-2A - Unit 1 are produced. The power-to-flow ratio reduces the power 2.3 Unit 2 2.3-2C - Unit 3 2.3-2

m

' level' trip and associated reactor power / reactor power-imbalance boundaries by 1.055%-Unit 1 for a 1% flow reduction.

1.07%-Unit 2 1.07%-Unit 3  ;

For Unit 1, the power-to-flow reduction ratio is 0.949, and for Units 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 of reactor coolant pump (s). The circuitry monitoring pump operational status 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 restrict the power level for the number of pumps in operation.

Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure set point is reached before the nuclear over-power trip set point. The trip setting limit shown in Figure 2.3-1A - Unit 1

,2.3-1B - Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure (2355 psig) has been 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 Tout-4539) setpoints shown in Figure 2.3-1A have been established to maintain 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 -4746)

(10.79 Tout -4579)

(10.79 Tout -4579)

Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant' 2.3-1B 2.3-1C temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F.

Reactor Building Pressure The high reactor building pressure trip sett ing limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a loss-of-coolant accident, even in the absence of a low reactor ;colant system pressure trip.

2.3-3

' Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1A. Two conditions are imposed when 2.3-1B 2.3-1C the bypass is used:

1. By administrative control the nuclear overpower trip set point must be reduced to a value 5 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 operation with part of the reactor protection system bypassed. This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The over power trip set point of 1 5.0% prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural .

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 only following reacter shutdown. After shutdown has occurred, reset the pump contact monitor power level trip setpoint to 55.0%.

B. Single Loop Operation Single loop operation is permitted only after the reactor has been tripped. After the pump contact monitor trip has occurred, the following j actions will' permit single loop operation: 1

1. Reset the pump contact monitor power level trip setpoint to 55.0%.

2. Trip one of the two protective channels receiving outlet temperature information from sensors in the Idle Loop.

3. Reset flux-flow setpoint to 0.949 (Unit 1) 0.961 (Unit 2) 0.961 (Unit 3)

REFERENCES (1) FSAR, Section 14.1.2.2 (4) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.7 (5) FSAR, Section 14.1.2.6 '

(3) FSAR, Section 14.1.2.8 2.3-4 l

-s THERMAL POWER LEVEL, *,

- - 120 110 (105.5) p - 100 +7 ,

g

"  ?

e l #

90 l (30,87)

(-40,84) ACCEPTABLE 4 PUMP l 80

~~

(70 0)

OPERAil0N l 70l I

l CCEPTABLE l

- 60l (30.60.3)

(-40,57.3) 3 & 4 PUMP OPERATION l  !(51.7) 50 l

I 1

g

-- 40l ACCEPTABLE l

(-40,30.2) 2,3 & 4 l Pl!MP .__ 30l l

OPERATION l l 20 s -l =.

i 10 ; 3 7 ii

,, I i ,, ii

! E! lE i  ;

-40 20 0 20 40 Reactor Paner Imaalance "

l l

PROTECTI'/E SYSTEM idAXIMUM ALLOWABLE SETPOI!!TS gg . OCONEE

, U1IT I NUCLEAR STATION lo Figure 2.3-2A

_,3_;

Table 2.3-1A Unit 1 ,

Reactor Protective System Trip Setting Limits Two Reactor One Reactor Four Reactor Three Reactor Coolant Pumps Coolant Pump Coolant Pumps Coolant Pumps Operating in A Operating in Operating Operating Single Loop Each Loop (Operating Power (operating Power (Operating Power (Operating Power Shutdown RPS Segment -100% Rated) -75% Nated) -461 Rated) -492 Rated) Bypass

1. Nuclear Power Max. 105.5 105.5 105.5 105.5 5.0 I3)

(1 Rated)

2. Nuclear Power Max. Based 1.055 t inws flow I .055 t imes flow 0.949 times flow 1.055 times flow Bypassed on Flow (2) and labalance, minus reduction minus reduction minus reduction minus reduction (Z Rated) due to imbalance due to imbalance due to imbalance due to imbalance

3. Nuclear Power Max. Based NA NA 551 (5)(6) 55% (5) Bypassed on Pump Honitors. (Z, Rated)

4. High Reactor Coolant 2355 2355 2355 2355 1720(4)

System Presbure, psig, Max.

5. Low Reactor Coolant 1800 1800 1800 1800 Bypassed p System Pressure, psig, Min.

6. Variable Low Reactor ( 11.14 T,, 470(9(I) ( 11.14 Tout - 4706 )I (11.14 Tout - 4706) (11.14 Tout - 4706 )(l} Bypassed Coolant System Pressure psig, Min.

7. Reactor Coolant Temp. 619 619 619 (6) 619 619 F., Ham.

t

8. High Reactor building 4 4 4 4 4 Pressure, psig, Max.

(1) T, g is in degrees Fahrenheit ("F). (5) Reactor power level trip set point produced by pump contact monitor reset to 55.01.

. (2) Reactor Coolant System Flow, I.

(6) Specification 3.1.8 applies. Trip one of the (3) Administratively controlled reduction set two protection channels receiving outlet temper-only during reactor shutdown, ature information from sensors in the idle loop.

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

i i

1 (3) Except as provided in specification 3.5.2.4.b, the reactor i shall be brought to the hot shutdown condition within four hours if the quadrant power tilt is not reduced to less than 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

b. If the quadrant tilt exceeds +3.41% Unit I and there is simultaneous 3.41% Unit 2 3.41% Unit 3 indication of a misaligned control rod per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60%

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

c. Except for physics test, if quadrant tilt exceeds 9.44% Unit 1, 9.44% Unit 2 9.44% Unit 3 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 range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each I percent tilt for the maximum tilt observed prior to shutdown.

e. Quadrant power tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.

3.5.2.5 Control Rod Positions

a. Technical Specification 3.1.3.5 does not prohibit the e ercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits ia Technical Specification 3.5.2.2.

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

c. Position limits are specified for regulating and axial power shaping control rods. Except for physics tests or exercising control rods, the regulating control rod insertien/ withdrawal limits are specified on figures 3.5.2-1A1, 3.5.2-1A2 and 3.5.2-1A3 (Unit 1); 3.5.2-1B1, 3.5.2-132 and 3.5.2-133 (Unit 2);

3.5.2-1C1, 3.5.2-1C2 and 3.5.2-IC3 (Unit 3) for four pump operation, and on figures 3.5.2-2A1, 3.5.2-2A2 and 3.5.2-2A3 (Unit 1); 3.5.2-231, 3.5.2-2B2 and 3.5.2-2B3 (Unit 2);

3.5.2-2C1, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) for two or three 3.5-8

im 7#

pucp operation. Also, excepting physics tests or exercising control rods, the axial power shaping control rod insertion / )

, , withdrawal limits are specified on figures 3.5.2-4A1, 3.5.2-4A2 and 3.5.2-4A3 (Unit 1). If the control rod position limits are exceeded, corrective measures shall be taken i==ediately to achieve an acceptable control rod position. An acceptable control rod position shall then be attained within two hours.

The minimum shutdcwn margin required by Specificatica 3.5.2.1 shall be maintained at all ti=es.

d. Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-1A1, 3.5.2-1A2 (Unit 1), 3.5.2-1B1, 3.5.2-1B2, and 3.5.2-1B3 (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-state rated power.

(2) The xenon reactivity shall be asymptotically approaching the value for operation at the power level cutoff.

3.5.2.6 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.

Except for physics tests, imbalance shall be maintained within the envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3. 5. 2-3 A3, 3.5. 3-3B1, l 3.5.2-3B2, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2, and 3.5.2-3C3. If the im-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 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.

3.5-9 .

.m .

Bases The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3A3, 3.5.2-3B1, 3.5.2-3B2, 3.5.2-333, 3.5.2-3C1, 3.5.2-3C2 and 3.5.2-3C3 is based on LOCA analyses which have defined the maximum linear heat rate (See Figure 3.5.2-5)-such that the maximum clad temperature will notl 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 effects

d. Hot rod manufacturing tolerance factors

e. Fuel rod bowing effects The 25% 1 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lover 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 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 0.65% ak/k at rated power. These values have been shown to be safe by l 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 positions limits at hot zero power. A single inserted control rod worth of 1.0% ak/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, therefore, less severe environmental con- i sequences than a 0.65% ak/k ejected rod worth at rated power. l

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

l 3.5-10

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 5.10% for Unit 3 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 axial imbalance monitoring in Specification 3.5.2.4 and 3.5.2.6, respectively, normally will be performed in the process 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 to prevent excessive power peaking by transient xenon. The xenon reactivity must be beyond the "undershoot" region and asymptotically approaching its equilibrium value at the power level cutoff.

REFERENCES 1

TSAR, Section 3.2.2.1.2 2

FSAR, Section 14.2.2.2 FSAR, SUPPLDfENI 9 B&W WEL DENSIFICATION REPORT BAW-1409 (UNIT 1)

RAW-1396 (UNIT 2)

BAW-1400 (UNIT 3) 5 0conee 1, Cycle 4 - Reload Report - BAW-1447, March 1977.

3.5-11

i OPERATION IN (94.102 (174.102) g ,,(225.9.102) 10 -

THIS REGION IS e POWER LEVEL 90 - (174.90) ,

25.9 0 RESTRICTE0 80 -

REGION (169.80) (230 9.80)

SHUTOOWN MARGIN 70 -

% (164.70) ( 235.9.70)

ll41T 60 -

(159,60) ( 240.1.60) 4 (300.60)

• 50 - (42.50)

(80 50) PERMISSIBLE R 10N 5 40 -

2 30 20 G, (0.15) (39.15) 10

' ' I I I I I ' ' ' '

0%

0 100 200 300 Rod inces % fitndraen 0 25 50 75 100 0 25 50 75 100 l  ! I I f I f f f f Group 5 Group 7 0 25 50 75 100 l  ?  ? f I Group 6 Roa inces is tne percentage sum of tne eitnarisal of Groups 5.6 ana 7 i

i ROD POSITION LIMITS FOR l FOUR-PUMP OPERATION FROM 0 i TO 100 (:10) EFPD, UNIT 1

'out ro OCONEE NUCLEAR STATION I 3.3-12

'M Figure 3.5.2-1A1 i

l l

(188.102) -

o(125.9.102) 100 -

RESTRICTED OPERATION IN THis poser LEVEL (225.9,90) 90 -  % (174.1.90) ,

REGION is NOT CUTOFF REGION ALLOWED (169.1,80) (230 9.80) 20 _

H (164.1.70) (235.9,70) 70 -

~

t240 9.60)

% 60 -

RESTRICTED REGION (159 1.60)

(300.60) 3 50 (126.50) o PERMISSIBLE OPERATING 40 REGION 30 -

20 -

(84,15) 10 -

(0.0)

' ' ' ' ' ' I I ' ' '

0 200 300 O 100 Rod index. 5 sitndrawn 0 25 50 75 100 0 25 50 75 100

! I i 1  ! 1 I t i  !

Group 7 Group 5 25 50 75 100 0

l l 1 1 I Group 6 Rod index is tne percentage sum of the sitndrawal of Gross 5.6 ana 7 ROD POSITION FOR FOUR-PUMP OPERATION FROM 100 (:10) TO 250 (:10) EFPD, UNIT 1 8

OCONEE NUCLEAR STATION Figure 3.5.2-1A2 3.5-13

s _.

" (251.8 102) 100 -

POWER LEVEL 90 - CUTOFF o (251.8 90i CPERATION IN THIS REGION IS NOT ALLOWED RESTRICTED S0 - (246.8.80)

REGION

- 70 - (241.8.70)

E M 60 - (236 8 BU) 50 - (127.50) d 40 -

}

SHUTOOWN MARGIN PERNISSl8LE OPER4 TING REGION 30 -

Llulf 20 -

(100.15) 10 ._

(0.0) 0 , , , , , , , , , , ,

0 100 200 300 Rod index, 5 Witndrawn J

0 25 50 7h 100 0 25 50 75 100 i t t t i i i i , i Group 5 Group 7 0 25 50 75 100 i t  ! I i Group 6 Roo inaen is tne ser:entage sum of the eitnarasal of Groups 5.6 and 7 i

R0D POSITION LIMITS FOR FOUR-PUMP OPERATION AFTER 250 (:10)

EFPD, UNIT 1 lount reets, OCONEE NUCLEAR STATION 3.5-13a W Figure 3.5.2-1A3

(94,102) (159.102) 100 3 9.1 ) (231.102)

~ OPERATION IN THIS 2 AND 3 PUMP RESTRICTED REGION IS NOT ALLOWED OPERATION REGION 0 (164,89)

1. 1TH 2 OR 3 PUMPS RESTRICTED 1 (236,89)

ERATION

= THIS REGIO 5 B0 -

RESTRICTED IN (241*76)

{ THl$ REGION (159,76) 70 - NINIMUM ( . 6)

$ SHUT 00nN LIMIT

$ 60 M

s 50 - t44,50)

= 40 -

PERMISSIBLE OPERATING REGION 2

"_ 30 -

Q 20 j Gs 0.15) 10 0 I I I I I t i e

!  ! t 0 100 200 300 Roa Index. 5 Witnaraon 0 25 50 75 100 0 25 50 75 100 I ' f f f I t I t i Group 5 Group 7 0 25 50 75 100 l t t I t Group 6 Roa inaen is the percentage sum of tne witnaranal of Groups 5.6 ana 7 R0D POSITION LIMITS FOR TWO-AND THREE-PUMP OPERATION FROM gg 0 TO 100 (-10) EFPD, UNIT 1 l

.ota m * % OCONEE NUCLEAR STATION D' Figure 3.5.2-2Al 3.5-18

., ~

(188.102) (231.102) 100 -

OPERATION IN THIS REG 10N RESTRICTED IS NOT ALLOWED REGION FOR 3 90 PUMP e -

(236.89)

OPERATION 2

20 -

SHUTOOWN MARGIN LIMIT (241.76)

$70 -

(300,76) 1 E SO -

.E t 50 -

(12S.50)

R 3 40 PERMISSIBLE OPERATING E REGION 30 O

E 20 a.

10 (84.0) 0 t I t t i t i i , , ,

0 100 200 300 Roa Index, *, Witnaraan 0 25 50 75 100 0 25 50 75 100 l I l l I I I I t l Group 5 Group 7 0 25 50 75 100 i t i  ! I Group 6 Roa inaex is tne percentage sum of the witnarasal of Groups 5.6 ana 7 R0D POSITION LIMITS FOR TWO-AND THREE-PUMP OPERATION FROM 100 (:10) TO 250 (:10), EFPD g UNIT 1

,ouni mw.t OCONEE NUC' EAR STATION 3.5-1Sa D' Figure .:. 5.2-2A2

, ..~, .-.

(236 8.102)

PERATION IN THIS REGION (201.102) i246 8.102) 100 IS NOT ALL0sEO WITH 2 OR 3 EGl0N PUMPS FOR 2 & 3

=

0 -

W pu,, (;4, g,gg, il

} 80 -

PERATION j RESTRICTED (236 8.76)

S IN THIS 7g _

R GION

[. SHUT 000N MARGIN LIMIT y 60 -

50 (127.50)

E 2 40 2

~

0 PERMISSIBLE OPERATING REGION 20 -

(100 15) 10 -

(0.0) 0  ! ' I I i i i t i l i 0 100 200 300 Rod inces. ", fitnaraan 0 25 53 75 100 0 25 50 75 100 l I f  ! I I , I i , i Group 5 Group 7 0 25 50 75 100 l 1 1 1  ! I Group 6 Roa races is tne percentage sum of tne eitncrasal of Grouus 5.6 ana 7 4

R00 POSITION LIMITS FOR TWO-AND THREE-PUMP OPERATION AFTER 250 (:10) EFPD, UNIT 1

.5 untron e, OCONEE NUCLEAR STATION 5

3.5-18b Figure 3.5.2-2A3 l

l

aM 9%

POWER,"~, of 2568 MWt RESTRICTED REGION 110

(-17.3,102) 100 -- f (9.0,102)

(-15.3,90) (8.8,90) 90 80 70 PERMISSIBLE OPERATING 60 RANGE 50 40 30 20 10 1

I I O i l l

-20 -

10 0 +10 +20 Core Imoalance, 5 l

l .

l l

OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM l 0 TO 100 (:10) EFPD, UNIT 1 e) OCONEE NUCLEAR STATION 3.5-21 Figure 3.5.2-3Al i

I

-. ~ . - . . .

t [ GL, POWER, ", OF 2568 MWt RESTRICTED REGION i10

(-18.3.102) ( +33,102 )

100

(-17.0,90) 90 (+3 8 90) 80 7C 60 50 40 30 20 10 I I I I

-20 - 10 0 +10 +20 Core Imualance, ",

l l

I OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 100 (:10) TO 250 (:10) EFPD, Ms

< UNIT 1 loutro*U, OCONEE NUCLEAR STATION g; Figure 3.5.2-3A2 3.5-21a l

POWER, 5 0F 2568 MWt 110 RESTRICTED REGION

(-23.1.102) f(15.8,102) 00

(-21.2,90) 90 (14.9,90)

PERMISSIBLE 80 OPERATING REGION 70 60 50 40 30 20 1 10  :

I I I I I l

-30 -20 -10 0 +10 +20 Core Imoalance, ",

OPERATIONAL POWER IMSALANCE ENVELOPE FOR OPERATION AFTER 250 (:10) EFPD, UNIT 1

.b Ast roara\

D, OCONEE NUCLEAR STATION 3.5-21b Figure 3.5.2-3A3 l

l

. .o .

l 100 _

1(37.0,102) 90 -

, ,(39.4,90) RESTRICTED REGION 80 - 5 (39.4,80) b 70 - (46.4.70) m M (90.4,60) 60 O

" (100,60)

50 -

J' -

40 PERMISSIBLE 30 20 -

10 -

l I I I I I I i 1 0

0 10 20 30 40 50 60 70 80 90 100 APSR $ Witndrawn APSR POSITION LIMITS FOR OPERATION FROM 0 T0 100 e 10) EFPD, UNIT 1 g (:CONEE

,ocu nan; O NUCLEAR STATION M Figure 3.5.2-4A1 -

3.5-23c i

i

(14.0,102) 100 -

9 (32,102)

[RESTRICTE0

, (34.6,90) RESTRICTED 90 -

(; , )

REGION 80 - (9.2,80) < (34.6,80) 70 -4 (3.6,70) (41.6,70)

_ (85.6,60) 60 * (3.6,60)  :

(100,60)

(0.60) o 50 PERMISSIBLE 40 OPERATING REGION 5

m 30 20 10 0 I I I I I I I 'l I O 10 20 30 40 50 60 70 80 90 100 APSR, ", Witnaraan l

l APSR POSITION LIMITS FOR

]

OPERATION FROM 100 (:10) l

, TO 250 (:10) EFPD, UNIT 1 l l ,'outrowit OCONEE NUCLEAR STATION 3.5-23d Wi Figure 3.5.2-4A2 l

l

s .-

(11.0,102) (32.0,102) RESTRICTED 100 ~

REGION 90 - , (8.6,90) ,(34.4,90) 80 -< .(6.2,80) o(34.4,80)

(6.1,70) ,

(34.4,70) 70 -

- (85.4,60)

E 60 w ,(6.1,60) _

(100,60) g (0.60) 50 PERMISSIBLE 40 d

OPERATING J' REGION 30 20 10 0 1 I I I I I I I I 10 20 30 40 50 60 70 80 90 100 O

APSR, 5 Witndrawn APSR POSITIOPl LIMITS FOR OPERATI0fl AFTER 250 (:10)

EFPD, UtlIT 1

[c *\ OCONEE NUCLEAR STATION 3.5-23e Y Figure 3.5.2 JA3

, ~.

. .a e 20-  : i i i i i i e i i i

, 18 -

t2 3

n

• 16 - -

a n

2 v

T

=

14 - -

2 Generic FAC BAW-10103 3_

s - _

s I 1; -

1C O 2 4 6 8 10 12 Axial Location of Peak Power From Bottom of Core, ft l

l l

1 LOCA-LIMITED MAXIMUM ALLOWABLE LINEAR HEAT

'ouu rown\

s N; OCONEE NUCLEAR STATION 3 . ).-24

, rigure 3. d. 2 .0 1

, _ . -_