ML19312B954
| ML19312B954 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/20/1974 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19312B949 | List: |
| References | |
| NUDOCS 7911270737 | |
| Download: ML19312B954 (21) | |
Text
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PROPOSED TEcnEICAL SPEcIncATION RE
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant i
system pressure, coolant temperature, and coolant flow during power operation of the plant.
Objective To maintain the 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 /
2.1-2B - Unit 2 2.1-2C - Unit 3 power-imbalance point is above the line for the specified flow, the sa'fety limit is exceeded.
Bases 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, there is a sharp reduction of the heat transfer coefficient, which would j
result in high cladding temperatures and the possibility of cladding failure.
Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the W-3 correlation.(1)
The W-3 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 corc 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 1.3.
A DNBR of 1.3 corresponds to 94.3% probability at a 99% confidence level that DNB will not eccur; this is considered a conservative margin to DNB for all 2.1-1
i 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 set points to correspond to the elevated location where the pressure is actually measured.
The curve presented in Figure 2.1-1A represents the conditions at which a 2.1-1B 2.1-1C minimum DNBR of 1.3 is predicted for the maximum possible thermal power (112%)
when four reactor coolant pumps are operating (minimum reactor coolant flow is 131.3 x 106 lbs/hr.) This curve is based on the following nuclear power peaking factors (2) with potential fuel densification effects; 2.67; F 1.78; F
= 1.50 F
=
=
H The design peaking combination results 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 more restrictive of two thermal 2.1-2B 2.1-2C limits and include the effects of potential fuel densification:
1.
The 1.3 DNBR limit produced by a nuclear power peaking factor of FN = 2.67 or the combination of the radial peak, axial peak and position oYtheaxialpeakthatyieldsnolessthana1.3DNBR.
2.
The combination of radial and axial peak that causes central fuel melting at the hc spot.
The limit is 20.15 kw/ft - Unit 1 19.8 kw/ft - Unit 2 19.8 kw/ft - Unit 3 Power peaking is not a directly observable quantity and therefore limits have l
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 l 2.1-2B 2.1-2C 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 2.1-1A is the most restrictive of all possible reactor 2.1-1B 2.1-1C coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.
2.1-3B 2.1-3C The curves of Figure 2.1-3B represent the conditions at which a minimum DNBR 2.1-3C 2.1-2
o'f 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the loccl quality at the point of minimum DNBR is equal to 15%, (3) whichever condition is more restrictive.
Using a local quality limit of 15% at the point of minimum DNBR as a basis for Curves 2 and 4 of Figure 2.1-3B is a conservative criterion even though the 2.1-3C quality of the exit is higher than the quality at the point of minimum DNBR.
The DNBR as calculated by the W-3 correlation continually increases from point i
of minimum DNBR, so that the exit DNBR is 1.7 or higher, depending on the pressure.
Extrapolation of the W-3 correlation beyond its published quality range of +15% is justified on the basis of experimental data. (4)
The maximum thermal power for three pump operation is 86% - Unit 2 86% - Unit 3 due to a power level trip produced by the flux-flow ratio 75% flow x 1.07 = 80%
1.07 = 80%
power plus the maximum calibration and instrument error.
The maximum thermal power for other coolant pump conditions are produced in a similar manner.
i For each curve of Figure 2.1-3B, a pressure-temperature point above and to the 2.1-3C lef t of the curve would result in a DN3R greater than 1.3 or a local quality at the point of minimum DNBR less then 15% for that particular reactor coolant pump situation.
The 1.3 DNBR curve for four pump operation is more restrictive i
than any other reactor coolant pump situation because any pressure / temperature I
point above and to the left of the four pump curve will be above and to the lef t of the other curves.
The safety limits presented for Oconee Unit 1 have been generated using BAW-2 critical heat flux (CHF) correlation and the actual measured flow rate at Oconee Unit 1.(5) This development is discussed in the Oconee 1, Cycle 2-Reload Report, reference.(6) These limits have been developed using the procedures previously discussed in the Bases with the exception of the CHF correlation and the measured flow rate.
The minimum DNBR for the BAW-2 correlation is 1.32 whichcorrespondstoa95%probabilityata99%confidencelevelthatDNgwill The flow rate utilized is 107.6% of the design flow (131.32 X 10 lbs/HR) occur.
based on four pump operation. (6) Because the four pump pressure-temperature restriction is known to 6e more limiting than the 3 and 2 pump combinations, only the four pump limit,has been shown on Figure 2.1-3A.
The maximum thermal power for three pump ope {ation for Unit 1 is 87% due to a power level trip produced by the flux-flow ratio 75% flow X 1.08 = 81% power plus the maximum calibration and instrument error.
REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) FSAR, Section 3.2.3.1.1.c (3) FSAR,'Section 3.2.3.1.1.k i
2.1-3
I (4) The following papera which were presented at the Winter Annual Meeting, ASME, November 18, 1969, during the "Two-phase Flow and Heat Transfer in Rod Bundles Symposium:"
(a) Wilson, et. al.
" Critical Heat Flux in Non-Uniform Heater Rod Bundles".
(b) Gellerstedt, et. al.
" Correlation of a dritical Heat Flux in a Bundle Cooled by Pressurized Water".
(5)
Correlation of critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, Mar. 1970.
(6) Oconee 1, Cycle 2 - Reload Report - BAW-1409, September, 1974.
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2500 2400
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a E
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2200
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5 g
2100 2000
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1900 560 580 600 620 640 660 Reactor Outlet Temperature, F 4
CORE PROTECTION SAFETY LIMITS UNIT 1 x
ocurim\\ OCONEE NUCLEAR STATION 2.1-4a ter Figure 2.1-1A
I f
Thermal F xer Level, 5 1
120 I
-100 1
t 2
" 80 (3AND4}
z
- 60
- - 40 20
-40
-20 0
+20
+40 Reactor Power imbalance, %
CURVE REACTORC00LAkTFLOW(LB/HR) 6 1
131.3 x 10 6
2 98.1 x 10 3
64.4 x 106 6
4 60.1 x 10 CORE PROTECTION SAFETY LIMITS UNIT 1
\\ OCONEE NUCLEAR STATI 2.1-7 Figure 2.1-2A
(
2500 2400
{
2300 a
a:
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2200 5
a 2100 2000
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1900 560 580 600 620 640 660 Reactor Outlet Temperature, F l
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CORE PROTECTION SAFETY LIMITS UNIT 1 2.1-10
) OCONEE NUCLEAR STATION Figure 2.1-3A
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2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.
Objective To provide automatic protective action to prevent any combination of process variables from exceeding a safety limit.
Specification The reacter protective system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1A - Unit 1 and 2.3-1B - Unit 2 Figure 2.3-2Al } Unit 1 2.3-2A2 2.3-2B - Unit 2 2.3-2C - Unit 3 The pump monitors shall produce a reactor trip for the following conditions:
Loss of two pumps and reactor power level is greater than 55% (0.0% for a.
Unit 1) of rated power.
b.
Loss of two pumps in one reactor coolant loop and reactor power level is greater than 0.0% of rated power.
(Reactor power level trip setpoint is reset to 55% of rated power for single loop operation. Reactor power level trip setpoint is reset to 55% for all modes of 2 pump operation for Unit 1.)
Loss o'f one or two pumps during two-pump operation.
c.
Bases The reactor protective system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits for protective system instrumentation are listed in Table 2.3-1A - Unit 1.
The safety analysis has been based upon these protective 2.3-1B - Unit 2 2.3-1C - Unit 3 system instrumentation trip set points plus calibration and instrumentation crrors.
Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
2.3-1 1
+'
l' During' normal plant operation with all reactor coolant pumps cperating, reactor trip is initiated when the reactor power level reaches 105.5% of rated power.
Adding to this the possible variation in trip setpoints due to calibration cnd instrument errors, the maximum actual power at which a trip would be actu-cted could be 112%, which is more conservative than the value used in the cafety analysis.(4)
Ovarpower Trip Based on Flow and Imbalance Tha power level trip set point produced by the reactor coolant system flow is bcsed on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has demonstrated 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.
Ths 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 1svel 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-i tzction for all modes of pump operation.
For every flow rate there is a maxi-rum permissible power level, and for every power level there is a minimum permissible low flow rate.
Typical power level and low flor rate combinations for the pump situations of Table 2.3-1A are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if power is 108% and reactor flow rate is 100%, or flow rate is 93% and power level is 100%.
2.
Trip would occur when three reactor coolant pumps are operating if power is 81.0% and reactor flow rate is 74.7% or flow rate is 69% and power level is 75%.
3.
Trip would occur when two reactor coolant pumps are operating in a single loop if power is 59% and the operating loop flow rate is 54.5% or flow rate is 43% 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 53% and reactor flow rate is 49.0% or flow rate is 45% and the power level is 49%.
For safety calculations the maximum calibration and instrumentation errors for tha power level trip were used.
Tha power-imbalance boundaries are established in order to prevent reactor th2rmal limits from being exceeded.
These thermal limits are either power p uking kw/ft limits or DNBR limits.
The reactor power imbalance (power in tha top half of core ninua power in the bottom half of core) reduces the power 1svel trip produced by the power-to-flow ratio such'that the boundaries of Figure 2.3-2A1 ) Unit 1 are produced.
The power-to-flow ratio reduces the power 2.3-2A2 2.3-2B - Unit 2 2.3-2C - Unit 3 2.3-2
I i
Shutdown Bypaan In order to provide for control rod drive tests, zero power physics testing, end startup proceduras, thersjm provision for bypassing certain segments of tha reactor protection system.
The reactor protection system segments which ccn be bypassed are shown in Table 2.3-1A.
Two conditions are imposed when 2.3-1B 2.3-lc the bypass is used:
1.
By administrative control the nuclear overpower trip set point must be reduced to a value ;E 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 opsration with part of the reactor protection system bypassed.
This high pressure trip met point is lower than the normal los pressure trip set point to that the reactor must be tripped before the bypass is initiated.
The over power trip set point of ;E 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 tha reactor coolant pumps were operating.
Two Pump Operation A.
Two Loop Operation Operation with one pump in each loop will be allowed only following reactor shutdown. After shutdown has occurred, the following actions will permit operation with one pump in each loop:
1.
Reset the pump contact monitor power level trip setpoint to 55.0%.
2.
(Unit 1) Reset the protective system taximum allowable setpoint as shown in Figure 2.3-2A2.
B.
Single Loop Operation Sindle loop operation is permitted only af ter the reactor has been tripped.
After the pump contact monitor trip has occurred, the following actions will permit single loop operation:
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 sen ors in the Idle Loop.
3.
(Unit 1) Reset the protective system maximum allowable setpoints as shown in Figure 2.3-2A2.
Tripping one of the two protective channels receiving outlet temperature information from the idle loop assures a protective system trip logic of one out of two.
REFERENCES (1) FSAR, Section 14.1.2.2 (5) FSAR, Section 14.1.2.6 (2) FSAR, Section 14.1.2.7 (3) FSAR, Section 14.1.2.8 (4) FSAR, Section 14.1.2.3 2.3-4
(
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leval. trip cnd czeccisted rarctor powar/rsectcr powar-ichslence boundaries by
'1.08% - Unit 1 for a 1% flow reduction.
1.07% - Unit 2 1.07% - Unit 3 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 overpower trip set point.
The trip setting limit shown in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2 2.3-lc - 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 (1985) psig and variable low pressure (13.77 Tout-618D trip (1800) psig (16.25 T
-7756)
(1800) psig (16.25T"h7756) setpoints shown in Figure 2.3-1A have been established to maintaSn 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 (13.77 Tout - 6221)
(16.25 T
-7796)
(16.25 T,
-7796)
Coolant Outlet Temperature The high yeactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 1 2.3-1B 12.3-1C 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 setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a loss-of-coolant accident, even le the absence of a low reactor coolant system pressure trip.
2.3-3
l 2400 2300
.5 E
J 2200 E
v, 0
Z 2100 8
a I;
a 2000 1900 i
i 540 560 580 600 620 640 Reactor Outlet Temperature, F i
PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SET POINTS l
l UNIT 1 2.3-5 ocur,'eis OCONEE NU. LEAR STATION o'
Figure 2.3-lh
i Pacer Level, 's 120 i
l FOUR PUNP 10 SET POINTS 0
-- 60 4
THREE PUMP SET POINTS
-- 40 20
-40
-20 0
20 40 Reactor Power imbalance, t I
l PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SET POINTS l
UNIT 1 n\\ OCONEE NUCLEAR STATIO 2.3-8 Figure 2.3-2Al
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1 Power Level, r, l
-- 120
-- 100 80 TWO PUMP 60 SET POINTS
-- 40 TWO PUMPS IN ONE LOOP DNE PUMP IN EACH LOOP
-- 20 i
R
-40
-20 0
20 40 i
Reactor Power imbalance, 5 l
l PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SET POINTS I
UNIT 1 Cevidem\\, OCONEE N 2.3-8a Figure 2.3-2A2
Tablo 2.3-1A Cnit 1 Reactor Protective System Trip Setting Limits Two Reactor One Reactor Tour Reactor Three Reactor Coolant Pumps Coolant Pump Coolant Pumps Coolant Pumps Operating in A Operating in Operating Operating Single Loop Each Loop (Operating Power (Ope 6ating Power (Operating Power (Operating Power Shutdown RPS Segment
-100% Rated)
-75% Rated)
-46% Rated)
-49% Rated)
Bypass 1.
Nuclear Power Max.
lu5.5 105.5 105.5 105.5 5.0(3)
-(% Rated) 2.
Nuclear Power Mat. Based 1.08 times flow 1.08 times flow 1.08 times flow 1.08 times flow Bypassed on Flow (2) and Imbalance, minus reduction minus redaction 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 55% (5)(6) 55% (5)
Bypassed on Pump Monitors, (%, Rated) 4.
High Reactor Coolant 2355 2355 2355 2355 1720(4)
System Pressure, psig, Max.
w
- y 3.
Low Reactor Coolant 1985 1985 1985 1985 Bypassed g
System Pressure, psig, Min.
6.
Variable Low Reactor (13.77 T
-6181)II) (13.77 T
- 6181)(1) (13.77 T
- 6181)(1) (13.77 T
- 6181)(1)
Bypassed out t
out Coolant System Pressure psig, Min.
7.
Reactor Coolant Temp.
619 619 619 (6) 619 619 F., Max.
8 High Reactor Building 4
4 4
4 4
Pressure, psig, Max.
. (1) T "" is in degrees Fahrenheit ( F).
(5) Reactor power level trip set point produced by pump contact monitor reset to 55.0%.
(2) Reactor Coolant System Flow, 2.
l' (6) Specification 3.1.8 applies. Trip one of the (3) Administratively controlled reduction set two protection channels receiving outlet te=per-only during reactor shutdown.
ature information from gensors in the idle loop.
(4) Automatically set when other_ segments of the RPS are bypassed.
1.
Rod inden is the percentage sum of the withdrawal of the operating groups.
2.
Tne withdrawal limits are modified af ter 25015 full power days of operation.
173 209 100 154 213 g4, Power Level
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Restricted Region 80 122 230
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Region an
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20 0
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0' 50 100 150 20J 250 300 Rod Index, S Withdrawal 25 50 75 100 E
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0,
,25
,50 75 100 sp7 0
25 50 75 100 Gp6 I
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1 1
Gp5 CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 4 PUMP OPERATION UNIT 1 3.5-12 ouu Ann OCONEE NUCLEAR STATION Figure 3.5.2-1A1
1.
Rod Index is the percentage sum of the withdrasal of tne operatent.
groups.
2.
T he withdrasal limits are in effect after 250 1 /; full poner days cf operation. (7he applicanle poner level cutoff is 100% power) 1 257.5 288.1 100 1
n E
80 Restricted Region E
E 135
[:
E 60 p
f (52% P) 4 Permissinle p
40 Operating Region i
20
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1 0
e i
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0 50 100 150 200 250 300 i
Rod index, S withdrawal 0
25 50 75 100 i
t t
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09 0
25 50 75 100 l
i I
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0 25 50 75 100 Gp6 i
1 1
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Gp5 CONTROL ROD GROUP WITHDRAWAL '.IMIT FOR 4 PUMP OPERATION l
UNIT 1 3.5-13 ouuronts OCONEE NUCLEAR STATION Figure 3.5.2-11.2 l
i 1.
Rod index is the percentage sum of the withdrawal of the operating groups. (The applicable power level cutoff is 100% power) 2 Pump Withdrawal Limit 230 ten 150 275 C
Restricted j
Region E
O 80 P
E Permissible ve
.[
Operating 64,'
f Region 60
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r
=
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E E
20 0
i i
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i 0
50 100 150 200 250 300 Rod index, $ withdrawal CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 3 AND 2 Pla!P OPERATION UNIT 1
\\ OCONEE NUCLEAR STATION 3.5-18 Figure 3.5.2-2A
(
Power, % of 2568 H t
__ 110
-20.4
+14.1 102%
__ 100 90
__ 80
- 30, 75 70
\\
s 60 50 l
-31.2,52
+28.1, 52 40
/
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i i
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-30
-20
-10 0
10 20 30 Core imbalance, %
OPERATIONAL POWER IFBALANCE ENVELOPE l
1 UNIT 1 ountrom) OCONEE NUCLEAR ST 3.5-21 Y'
Figure 3.5.2-3A
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20
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Generic FAC 18 Results (BAW-10091)
\\ Batch 2 & 3 Fuel J
ce 14 0
x E
3 12 2
m 10 0
2 4
6 8
10 12 Distance from inlet, it I
LOCA LIMITED PAXIMUM ALLOWABLE LINEAR HEAT RATE
\\ OCONEE NUCLEAR STATION 3.5-24 Figure 3.5.2-4
.