ML19317F320
| ML19317F320 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/06/1977 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19317F317 | List: |
| References | |
| NUDOCS 8001100695 | |
| Download: ML19317F320 (26) | |
Text
~.
(T)
I I
Bases - Unit 3 The safety limits presented for Oconee Unit 3 have been generated using BAW-2 critical heat flux correlation (l) and the Reactor Coolant System flow rate of 106.5 percent of the design flow (131.32 x 106 lbs/hr for four-pump operation).
The flow rate utilized is conservative compared to the actual measured flow rate.(2)
To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prever.t overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature.
The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB).
At this point, there' is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure.
Although DN3 is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be relat,ed to DNB through the use of the BAW-2 correlation (l).
The BAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the 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 confi-dance level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions.
The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.
The difference in these two pressures is nominally 45 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-lc represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 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:
N y
= 2.67; F I
= 1.50.
The design peaking AH z
combination 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.
1.
The 1.30 DNBR limit produced by a nuclear peaking factor of F
= 2.67 or the combination of the radial peak, axial peak and position o the axial peak that yields no less than a 1.30 DNBR.
2.1-3e
'8001100( 7
s 2.
The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.15 kw/f t for Unit 3.
d 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 in each loop, respectively.
The maximum 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 maximum calibration and instrument error.
i 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 lef t of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation.
The curve of Figure 2.'l-1C 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 Bundle Cooled by Pressurized Water, BAW-10000, March 1970.
(2) Oconee 3, Cycle 3 - Reload Report - BAW-1453, August, 1977.
l 0
n 2.1-3d
.u
2400 2200 ACCEPTABLE OPERATION m.
5 5
2000 n.
3 UNACCEPTABLE o
OPERATION a,
1800 I
1600 560 580 600 620 640 Reactor Coolant Outlet Temperature,F i
CORE PROTECTION SAFETY LIMITS UNIT 3 OCONEE NUCLEAR STATION
?.1-6 Figure 2.1-1C
J THEHilAL power LEVEL, s
- 120
-37,112 37,112
-- 110 O
- 100 ACCEPTABLE 4-PUMP OPERATION 9
-37,85.3 37,85.3
-- 80
-52,80 49.2,80 g
ACCEPTABLE 3 & 4
-- 10 PUMP OPERATION
-37,58.2 60 37,58.2 52,53.3 50
-- 40 ACCEPTABLE 2,3 & 4
-- 30 PUMP OPERATION
-52,26.2 20 49.2,26.2 10 I
i i
i i
i i
1 i
I
-60
-40
-20 0
20 40 60 i
1 Reactor Power imbalance, %
Curve Reactor Coolant Flow, gpm_
1 374,880 (100%)*
2 280,035 (74.7%)
3 183,690 (49.0%)
CORE PROTECTION SAFETY LIMITS UNIT 3
- 106.5% or ftrst-core design flow 2.1-9 i OCONEE NUCLEAR STATION Figure 2.1-2C L
2400
/
ACCEPTABLE OPERATION 2200 a
a a
E 2000
/
2 3
5
- 1. f
/
O 1800 1600 i
i i
1 560 580 600 620 640 Reactor Coolant Outlet Temperature,F
- Power, Pumps Type of Curve Coolant Flow, gpm Operating Lim i t_ _
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.
CORE PROTECTION SAFETY LIMITS UNIT 3
} OCONEE NUCLEAR STATION 2.1-12 Figure 2.1-3C
LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 2.3 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 reactor protective system trip setting limits and the permissible bypasses I and for the instrument channels shall be as stated in Table 2.3-1A - Unit 2.3-1B - Unit 2 2.3-1C - Unit 3 Figure 2.3-2A - Unit 1 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% of rated a.
power.
b.
Loss of two pumps in one reactor coolant loop and reactor power level is greater than 0.0% of rated power. (Power /RC pump trip setpoint is reset l
to 55% for 2 pump operation.)
Loss of one or two pumps during two-pump operation.
)
c.
i 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 errors.
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 i
detected by pressure and temperature measurements.
9 2.3-1 u
level trip and associated reactor power / reactor power-imbalance boundaries by 1.0 55%
for a 1% flow reduction.
Pump Monitors The pump monitors prevent tripping the reactor due to the loss of reactor coolantthe minimum core DNBR fro pump (s).
The circuitry monitoring pump operational status provides redundant by tripping the reactor on a signal diverse f rom that of the power-to-flow trip protection for DNB ratio.
The pump monitors also restrict pumps in operation.
the power level for the number of Reactor Coolant System Pressure During a startup accident power, the system high pressure set pointfrom low power or a slow rod withdrawal from 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 for high reactor coolant 2.3-1C - Unit 3 maintain the system pressure below the safety limitsystem pressure (2355 psig) has bee (1)
(2750 psig) for any design transient.
The low pressure (1800) psig and variable low pressure (11.14 T
-4706) trip (1800) psig out (1800) psig (11.14 Tout-4706)
(11.r4 Tout-4706) setpoints shown in Figure 2.3-1A have been established to maintain the DNB 2.3-1B 2.3-1C rctio greater than or equal to 1.3 for those design accidents that a pressure reduction. (2,3) result in Due to the calibration and instrumentation errors the safety analysis used a v:riable low reactor coolant system pressure trip value of (11.14 T
-4746)
(11.14 Tout -4746)
Coolant Outlet Temperature (11.14 Tout -4746) l 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 tteperatures in the operating range.
crrors, the safety analysis used a trip set point of 620 F.Due to calibration and instrume R^ actor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides p;sitive 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 coolant cystem pressure trip.
2.3-3
4 Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, cnd startup procedures, there is provision for bypassing certain segments of the reactor protection system.
The reactor protection system segments which ecn be bypassed are shown in Table 2.3-1A.
Two conditions are imposed when 2.3-1B 2.3-1C the bypass is used:
By administrative control the nuclear overpower trip se,t point must be 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 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 go 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 i
Operation with one pump in each loop will be allowed only following reactor shutdown.
After shutdown has occurred, reset the pump contact monitor power level trip setpoint to 55.0%.
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
i o
2400 T = 619*F 7
P = 2355 psig 2300 -
i
.T 2200 ACCEPTABLE g
j OPERATION g
s N
5 2l00 E
2 UNACCEPTABLE j
OPERATION g
v 2000 -
E
%o 6
1900 g
P = 1800 psig 1800 1 = 584F i
l I
l 540 560 580 600 620 640 Reactor Outlet Temperature, F
PROTECTIVE SYSTEM MAXIMUM ALLOWhBLE SETPOINTS UNIT 3 2.3-7
- \\
w; OCONEE NUCLEAR STA Figure 2.3-lC
THERMAL POWER LEVEL, 5
-. 120 UNACCEPTABLE OPERATION 26.1,105.5
~~
25.0,105.5
+
ACCEPTABLE 100
!e
!4PUNP
- y g
l
- e, l OPERATION 90 l26.1,78.8 25.0, 78.8
-- 80
-40.1,73.5 36.9,73.5 l ACCEPTABLE 3 & 4 PUNP 70 OPERATION I
60 l26.1,51.7 l
25.0,51.7 i 50 36.9,46.8
-40.1,46.8 ACCEPTABLE 2,3, & 4 PUMP 40 OPERATION l
30 6.9,19.7
-40.1,19.7 20 2
o!
M T
[g 10 d
,, I ii o
d E
i i
i n
i e
if n
i i
i 60 -50
-40
-30
-20
-10 0
10 20 30 40 50 60 Power Imnalance, 5 PROTECTIVE SYSTEM MAXIMUM ALLOWABLF-SETPOINTS UNIT 3 2.3-10 s
nai tcars, OCONEE NUCLEAR STATION Figure 2.3-2C
Tchla 2.3-1C Unit 3 Reactor Protective System Trip Setting Limits One Reactor Four Reactor Three Reactor Coolant Pump Coolant Pumps Coolant Pumps Operating in Operating Operating Each Loop (Operating Power (Operating Power (Operating Shutdown
-100% Rated).
-75% Rated)
-49% Rated)
Bypass RPS Segment 5.0( }
1.
Nuclear Power Max.
105.5 105.5 105.5
(% 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
- 4
- 3.. Nuclear Power Max. Based NA N/A 55%
Bypassed y
on Pump Monitors, (% Rated) 1720(
2355 2355 2355 4
High Reactor Coolant System Pressure, psig Max.
5.
Low Reactor Coolant 1800 1800 1800 Bypassed System Pressure, psig, Min.
-4706)( (11.14 T,- 4706)( }
(11.14 T,
-4706)(1)
Bypassed 6.
Variable Low Reactor (11.14 T Coolant System Pressure psig, Min.
7.
Reactor Coolant Temp.
619 619 619 619 F., Max.
4 4
4 4
8.
High Reactor Building Pressure, psig, Max.
(1) T is in degrees Fahrenheit ( F).
(2) R0 actor Coolant System Flow, %.
(3) Administratively controlled reduction set only during reactor shutdown.
(4) Automatically set when other segments of the RPS are bypassed.
.. o,
pump opersti.
Also, excepting physics test 9r exercising rontrol rods, tha axial power shaping control ad insertion /
withdrawal limito are specified on figures 3.5.2-4A1, 3.5.2-4A2 and 3.5.2-4A3 (Unit 1), 3.5.2-4B1, 3.5.2-4B2, and 3.5.2-4B3 i
(Unit 2), and 3. 5.2-4C1, 3. 5.2-4c2, and 3.5. 2-4C3 (Unit 3).
l If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position.
An acceptable control rod position shall then be attained within two hours. The minimum shutdown margin required by Specification 3.5.2.1 shall be maintained at all
- times, d.
Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-1A1, 3.5.2-1A2 (Unit 1), 3.5.2-151, 3.5.2-182, and 3.5.2-1B3 (Unit 2), and 3.5.2-Icl, 3.5.2-Ic2, 3.5.2-Ic3 (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. worth has passed its final maximum or minimum peak during its approach to its equilibrium 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, 3.5.2-3A3, 3. 5.2-3B1 l
3.5.2-352, 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 or his designated alternate.
)
1 3.5-9 e
[
4 Bases _
3.5.2-3A2, 3.5.2-3A3, The power-imbalance envelope defined in Figures 3.5.2-3 is based on (see Figure LOCA analyses which have defined the maximum linear heat rate l
3.5.2-5) such that the maximum clad temperature will not ex Acceptance Criteria.
rod position, or imbelance be outside their the indicated quadrant tilt, Operation in a situation that would cause the Finalhighly specified boundary. Acceptance Criteria to be approached should a LOCA occ tilt, rod position, because all of the power distribution parameters (quadrantly all other and imbalance) must be at their limits while simultaneousuncertainty Conservatism engineering ano f
is introduced by application o :
Nuclear uncertainty factors
}
a.
Thermal calibration (Units 1 and 2 only) b.
Fuel densification power spike f actors rod manufacturing tolerance factors l
c.
Hot d.
Fuel rod bowing power spike factors e.
5% overlap between successive control rod groups is allowed since t of the stroke.
The 25%
the worth of a rod is lower at the upper and lower parControl Function Group Safety 1
Safety 2
Safety 3
Safety 4
Regulating 5
i Regulating 6
APSR (axial power shaping bank) 7 l
8 following three The rod position limits are based on the most limiting of h.
d by the Therefore, compliance with the ECCS power peaking criterion criteria:
l ip at any rod position limits. position limits, provides for achieving hot shutdown time, assuming the highest worth control rod thatThe rod position limi 0.65%
o position (1).
will not contain single rod wortha greater thanThese values have been shown to be safe out at rated power.
ident.
by the safety analysis (2, 3, 4, 5) of the hypothetical rod ejection a ak/k d by the A maximum single inserted control rod worth of 1.0% Ak/k is alloweA si i t rod position limits at bot zero power.1.0% ak/k at beginning-of-life, l consequences than peak thermal power and, therefore, lass severe environmentaejected rod O.65% Ak/k excore' detectors or
- Actual operating limits depend on whether or not incore The method used
~
i rrors.
are used and their respective instrument calibrat on erating procedures.
to define the operating limits is defined in plant ope
)
~
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 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 witndrawal 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 po.wer peaking by transient xenon. The xenon reactivity must be beyond its final maximum or minimum peak and approaching its equili-br'iuk value at the power level cutoff.
REFERENCES FSAR, Section 3.2.2.1.2 2FSAR, Section 14.2.2.2 FSAR, SUPPLEMENT 9 B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1)
BAW-1396 (UNIT 2)
RAW-1400 (UNIT 3) 5OCONEE UNIT 1, CYCLE 4 RELOAD REPORT, BAW-1447, March, 1977 Section 7.11 3.5-11
113.102 174.1,102 i, q,
32.0,102 100 RESTRICTED REGION OPERATION IN THIS
~32.0,90 y _
REGION l$ NOT 174.1,90 i
ALLOWE0 OWER LEVEL CUTOFF DS 161.2,80 251.4,80 SHUT 00tW 150.70 10 ungGIN 300,70 LIEli
[g N
~
80,60 50 36,50 48
~
PERNISSistE OPERATING REGION 30 20
,15 la 0,0 i
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0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Roa Index, 5 Iltnaraan 0
25 50 15 100 0
25 50 15 100 i
i i
i i
i i
i i
j Group 5 Group 7 0
25 50 75 100 i
1 i
l I
Group 6 R0D POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 TO 100 (+ 10) EFPD UNIT 3 g',) OCONEE NUCLEAR STATION e
3.5-16 Figure 3.5.2-101
ti232.0,102 174.!,102 100 -
RESTRICTED REGION b l74.1.90 i
32.0,90 90 -
OPERATION IN THl$
N POWER LEVEL REGION IS NOT ALLOWED CU10FF 161.2.80 mg 70 300,70 SHUT 005N MARGIN LIMll j
80
~
50 108,50 k
E PERMIS$18LE OPERAllHG 30 REGION 10,15 0,15 0,I.2 10 0,9 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Innet, 5 Witharaan 0
25 50 75 100 0
25 50 75 100 i
i 3
i i
i i
Group 5 Group 7 0
25 50 75 100 t
l I
I I
Group 6 R0D POSITION LIMITS FOR FOUR PUMP OPERATION FROM 100 (~+ 10)
EFPD TO 235 (+ 10) EFPD UNIT 3
@' OCONEE NUCLEAR Figure 3.5.2-102 3.5-16a
251.4,102 208,102 o
POIER 251.4,90
,e 90 -
OPERAil0H IN THIS REGION 15 NOT L HL CUT 0FF ALLOWED 241.7,80 00 -.
SHUTDOWN NARGIN Liitti
~
3 60 PERillSSIBLE OPERATING
~
REGION Eg 40 30 20 180.15 110,15 0,5.5 10
,0, 0 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 i
Rod index, 'S Withdrawn 0
25 50 75 100 0
25 50 75 100 e
i i
a i
i i
e Group 7 Group 5 0
25 50 75 i00 f
i 1
1 1
Group 6 R0D POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 235
(+ 10) EFPD UNIT 3 OCONEE NUCLEAR STATION nur, oat 3.5-17 Figure 3.5.2-1C3 i
x l
113.102 161.2.102 251.4.102 100 -
l RESTRICIE0 90 -
OPERAil0N IN THIS REGION FOR 3 150.89 PUMP REGION IS NOT OPERATION 300,85 g
ALLOWE0 3
0.75 5 10 I
SHUTOOWN NARGIN
= 60 j
Lluli E 50 36.50 PERMISSIBLE OPERATING REGION m
40 a
E 3 30
=
~
20 J
15 5 10 0.0 l
t t
t t
t f
1 1
1 I
I i
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Red inasx. 5 Withdrawn 0
25 50 75 100 0
25 50 75 IOC i
i t
i i
t i
i Group 5 Group 7 0
25 50 75 100 i
t t
t t
Group 6 l
ROD POSITION LIMITS FOR TWO-AND THREE-PllMP OPERATION FROM 0 TO 100 (i 10) EFPD UNIT 3 k OCONEE NUCLEAR STATION Figure 3.5.2-2Cl i
5 174.l.10 251.4,102 90 -
OPERATION IN THIS REGION RESTRICTED FOR 3 PUMP OPERATION g
15 NOT ALLOWE0 300,89 E 80
", 70 E
E0 SHUT 00tN MARGIN LIMIT PERMISSIBLE OPERATING REGION a
50 108,50 E
40
- 30 I 20 s'
70.15 to ~0,s.2 ESTRICTED FOR
+
2 & 3 PUMP OPERATION O,0 i
O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index, 5 tithdrawn 0
25 50 15 100 0
25 50 15 100 1
I I
v i
i 1
I i
Group 5 Group 7 0
25 50 75 100 t
f f
1 1
Group 6 l
R00 POSITION LIMITS FOR TWO-AND THREE-PUMP OPERATION FROM 100 (i 10) TO 235 (i 10) EFPD UNIT 3 3.5-20a i OCONEE NUCLEAR STATION Figure 3.5.2-2C2
RESTRICTED FOR 3 PUMP OPERAil0N WM 208,102 100 -
OPERAll0N IN THIS REGION lt NOT ALLOWED 90 -
196,92 7
80 5 10 SHUTOOWN NARGlN LIMii u
EO g
E 50 142,50 PERMISSIBLE OPERATING REGION
[o 40 2
E 30 2
20 100.15 "a 10 0,5.9 ESTRICTED FOR 2 & 3 PUMP OPERAil0N a-
=
i i
e i
t i
i t
i 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300
)
Rod Index, 5 Withdrawn 0
25 50 15 100 0
25 50 15 100 i
3 i
j Group 5 g,,,,
7 i
0 25 50 75 100 J
t i
i Group 6 R0D POSITION LIMITS FOR TWO-AND THREE-PUMP OPERATION I
AFTER 235 (+ 10) EFPD UNIT 3 OCONEE NUCLEAR STATION 8
3.5-20b
,,u ra ii.
W Figure 3.5.2-2C3
/
i t
Power, % of 2568'MWt RESTRICTED REGION
^ ' '
-13.73,102
- 100
^
-23.54,90 90 11.35,90 80
)12.63,80
-26.66,60 (
70 60 50 40 PERMISSIBLE
~ ~
OPERATING REGION 20 10 t
i I
I l
I 1
I e
i
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power imbalance, 5 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 100 (+ 10) EFPD UNIT 3 3.5-23 OCONEE NUCLEAR STATION Figure 3.5.2-3C1 P
1 1
i Power, 5 of 2568 M t RESTRICTED REGION
-25.57,102 g:
-100
-24.65,90i D
- - 90 1 11.35,90
-26,11,80 <
80 4 ) 12:63,80 70 l
-. 60 PERMISSIBLE
- 50 OPERATING REGION
- - 40 30 20 10 t
i i
i t
i t
i i
J
-50 40
-30
-20
-10 0
10 20 30 40 50 Axial Power Imnalance, 5 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 100 (+) 10 EFPD TO 235 _(+ 10)
EFPD (NIT 3 OCONEE NUCLEAR STATION Figure 3.5.2-3C2
f Power, 5 of 2568 MWt RESTRICTED REGION
-27.19,102 (p 100 90
() 18.42,90
-29.94,90 ()
i
-29.94,80 ()
- - 30
( ) 19.38,80 10 PERMISSIBLE 60 OPERAT ING REGION 50 40
. 30
-- 20 10 e
i i
e i
i i
i i
i 50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power imcalance, %
OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 235 (+ 10) EFPD UNIT 3 OCONEE NUCLEAR STATION 3.5-23b Figure 3.5.2-3C3
. i 100 S.6,102 33.3,102 RESTRICTED REGION 90 0,90 38.4,90 f
STRICTE0 64.4,80 80 REGION 70 100,70 E
60 3
m 50 a
c PERillSSIBLE 40
=g OPERATING REGION 30 20 10 0
O 10 20 30 40 50
'0 70 80 90 100 6
APSR, 5 Withdrawn I
APSR POSITION LIMITS FOR OPERATION FROM 0 TO 100 + 10 EFPD UNIT 3 3.5-231 out rowie OCONEE NUCLEAR STATION M
Figure 3._5.2-4C1
Figure 8-16.
APSR Position Limits for Operation From 100 ! 10 to 235 1 10 EFPD -- Oconee 3, Cycle 3 j
t
)
l 44.8,102 100 RESTRICTED REGION 51.4,90 90 64.4,80 80 70 100,70 5
60 3N 50 e
d 40
=
d' PERMISSIBLE 30 OPERATING REGION 20 10 I
I I
e i
0 0
10 20 30 40 50 60 70 80 90 100 APSR, 5 Withdrawn APSR POSITION LIMITS FOR OPERATION FROM 100 + 10 TO 235 i 10 EFPD UNIT 3 3.5-23j suryts OCONEE NUCLEAR STATION 3.5.2-4C2
i i
i a
APSR Position Limits for Operation After i
i Figure 8-17.
10 EFPD - Oconee 3, Cycle 3 i
235 f
RESTRICTED 36.3,102 100 -
REGION 51.4.90 90 -
l 64.4,80 80 100,70 70 r 5*
60 3
N PERMISSIBLE OPERATING REGION 50 e
40 a
n.
30 20 10 i
i t
i i
i 0
L_
0 10 20 30 40 50 60 70 80 90 100 APSR, % Withdrawn APSR POSITION LIMITS FOR OPERATION AFTER 235 + 10 EFPD UNIT 3 OCONEE M N M O M 3.5-23k
"'Pa Figure 3.5.2-4C3 L
l-m I I
s
.