ML19312C009
| ML19312C009 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/09/1975 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19312B997 | List: |
| References | |
| NUDOCS 7911270783 | |
| Download: ML19312C009 (25) | |
Text
'3 5
At:achment 1 Proposed Technical Specification Revisions Replacement Pages l
7911270 2h3
<m A
')
3.1.3 Minimum Conditions fer Criticelity Specification 3.1.3.1 The reactor coolant temperature shall be above 525 F except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply.
3.1.3.2 Reactor coolant temperature shall be above DTT + 10 F.
3.1. ' 3
'a the reactor coolant temperature is below the minimum temperature sp;;Ified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply, the reactor shall be suberitical by an amount equal to or greater than the calculated reactivity insertion due to depressurization.
3.1.3.4 The reactor shall be maintained suberitical by at least 1%Ak/k until a steam bubble is formed and a water level between 80 and 396 inches is established in the pressurizer.
3.1.3.5 Except for physics tests and as limited by 3.5.2.1, cafety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality.
The regulating rods shall then be positioned within their position limits defined by Specification 3.5.2.5 prior to deboration.
Bases At the beginning of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly positive at operating temperatures with the operating configuration of control rods.(1)
Calculations show that above 525 F, the con-sequences are acceptable.
Since the moderator temperature coefficient at lower temperatures will be less negative or more positive than at operating temperature,(2) startup and operation of the reactor when reactor coolant temperature is less than 525 F is prohibited except where necessary for low power physics tests.
The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1%ak/k.
During physics tests, special operating precautions will be taken.
In addition, the strong negative Doppler coefficient (1) and the small integrated Ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.
The requirement that the reactor is not to be made critical below DTT + 10 F provides increased assurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NDTT of the primary coolant system.
Heatup to this temperature will be accomplished by operating the reactor coolant pumps.
If the shutdown margin required by Specification 3.5.2 is maintained, there i
is no possibility of an accidental criticality as a result of a decrease of coolant pressure.
i 3.1-8
\\
,y s
)
The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1% suberitical will assure that the reactor coolant system cannot b cpe solid in the event of a rod withdrawal accident a start-up accident. 3e or The requirement that the safety rod groups be fully withdrawn before criti-cality ensures shutdown capability during startup.
This does not prohibit rod latch confirmation, i.e., withdrawal by group to a maximum of 3 inches withdrawn of all seven groups prior to safety rod withdrawal.
The requirement for regulating rods being within their rod position limits ensures that the shutdown margin and ejected rod criteria at hot zero power are not violated.
' REFERENCES (1)- FSAR, Section 3 (2) FSAR, Section 3.2.2.1.4 (3) FSAR, Supplement 3, Anewer 14.4.1 3.1-9
T I
3.1.7 Moderator Temperature Coefficient of Reactivity Specification The moderator temperature coefficient shall not be positive at power levels above 95 percent of rated power.
Bases A non-positive moderator coefficient at power levels above 95% of rated power is specified such that the maximum clad temperatures will not exceed the Final Ac-ceptance Criteria based on LOCA analyses.
Below 95% of rated power the Final Acceptance Criteria will not be exceeded with a positive moderator temperature coefficient of +0.9 x 10-4 Ak/k/ F corrected to 95% rated power. All other ac-cident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0.9 x 10-4 Ak/k/ F.
The moderator coefficient is expected to be zero or negative prior to completion of startup tests.
When the hot zero power value is corrected to obtain the hot full power value, the following corrections will be applied.
A.
Uncertainty in isothermal measurement The measured moderator temperature coefficient will contain uncertainty on the account of the following:
1.
9 2*F in the AT of the base and perturbed conditions.
2.
Uncertainty in the reactivity measurement of +0_.1 x 10-4 Ak/k.
l Proper corrections will be added for the above conditions to result in a conservative moderator coefficient.
B.
Doppler coefficient at hot zero power During the isothermal =oderator coefficient measurement at hot zero power, the fuel temperature will increase by the same amount as the moderator.
The measured temperature coefficient must be increased by 0.16 x 10-4(ak/k)/*F to obtain a pure moderator temperature coefficient.
C.
Moderator temperature change The hot zero power measurement must be reduced by.09 x 10-4 (ak/k)/*F. This corrects for the difference in water temperature at zero power (532*F) and 15% power (580*F) and for the increased fuel temperature effects at 15% power. Above this power, the average moderator temperature remains 580'F.
However, the co-efficient, g, must also be adjusted for the interaction of an average moderator temperature with increased fuel temperatures.
This correction is
.001 x 10-4 a % a% power.
It adjusts the 15%
/
power am to the moderator coefficient at any power level above 15%
power.
For example, to correct to 100% power, a is adjusted by m
(.001 x 10-4) (85%), which is
.085 x 10-44%.
3.1-17
3.5.2 Control Rod Group and Power Distribution Limits Applicability This specification applies to power distribution and operation of control rods during power operation.
Objective To assure an acceptable core power distribution during power operation, to set a l'mit on potential reactivity insertion from a hypothetical control rod ejcetion, and to assure core subcriticality after a reactor trip.
Specification 3.5.2.1 The available shutdown margin shall be not less than 1% ak/k with the highest worth control rod fully withdrawn.
3.5.2.2 Operation with inoperable rods:
l a.
If a control rod is misaligned with its group average by more i
than an indicated nine (9) inches, the rod shall be declared i
The rod with the greatest misalignment shall be evaluated first.
The position of a rod declared inoperable due to misalignment shall not be included in computing i
the average position of the group for determining the j
operability of rods with lesser misalignments.
b.
If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.
c.
If a control rod cannot meet the requirements of Specification 4.7.1, the rod shall be declared inoperable.
d.
If a control rod is found to be improperly programmed per Specification 4.7.2, the rod shall be declared inoperable until properly programmed.
e.
Operation with more than one inoperable rod in the safety or regulating rod groups shall not be permitted.
f.
If a control rod in the regulatin,g or safety rod groups is declared inoperable in the withdrawn position, an evaluation shall be initiated immediately to verify the existance of 1%
ok/k hot shutdown margin.
Boration may be initiated either to the worth of the inoperable rod or until the regulating and transient rod groups are fully withdrawn, whichever occurs first.
Simultaneously, a program of exercising the remaining regulating and. safety rods shall be initiated to verify operability.
3.5-6 7/19/74
')
i g.
If within one (1) hour of determination of an inoperable rod, it is not determined that a 1%Ak/k hot shutdown margin ex'ists combining the worth of the inoperable rod with each of the other rods, the reactor-shall be brought to the hot standby condition until this margin is established.
h.
Following the determination of an inoperable rod, all rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and ex'ercised weekly until 'the rod problem is solved.
1.
If a control rod in the regulating or safety rod groups is declared inoperable, power shall be reduced to 60 percent of the thermal power allowable for the reactor coolant pump com-bination.
j.
If a control rod in the regulating or axial power shaping groups is declared inoperable, operation above 60 percent of rated power may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 3.5.2.2.a and the withdrawal limits of Specification 3.5.2.5.c.
3.5.2.3 The worths of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the
, control rod position limits defined in Specification 3.5.2.5.
3.5.2.4 Quadrant Power Tilt a.
Whenever the quadrant power tilt exceeds 4 percent, except for, physics tests, the quadrant tilt shall be reduced to less than 4 percent within two hours or the following actions shall be taken:
(1) If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of full power for each I percent tilt in excess of 4 percent below the power level cutoff (see Figures 3.5.2-1A1, 3.5.2-1B1, 3.5.2-1B2, 3.5.2-1B3, 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3).
(2) If less than four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of i
l full power for each I percent tilt below the power allowable for the reactor coolant pump combination as defined by Specification 2.3.
(3) Except as provided in 3.5.2.4.b, the reactor shall be brought i
to the hot shutdown condition within-four hours if the quadrant tilt is not. reduced to less than 4 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. If the quadrant tilt exceeds 4 percent and there is simultaneous indication of a misaligned control rod per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60 percent of the thermal power allowable for the reactor coolant 3.5-7 l
l l
O
/
T 9
pump combination.
r c.
Except for physics tests, if quadrant tilt exceeds 9 percent, a controlled shutdown shall be initiated immediately and the reactor shall be brought to the hot shutdown condition within four hours.
4.
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 1 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 exercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.
b.
Operating rod group overlap shall be 25% + 5% between two sequential groups, except for physics tests.
c.
Except for physics tests or exercising control rods, the control rod withdrawal limits are specified on Figures 3.5.2-1A1 and 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, and 3.5.2-1C3 (Unit 3) for four pump operation and on Figures 3.5.2-2A (Unit 1), 3.5.2-2B (Unit 2), and 3.5.2-2C (Unit 3) for three or two pump operation.
If the control rod position limits are exceeded, corrective measures shall be taken immediately to l
achieve an acceptable control rod position.
Acceptable control rod j
positions 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 3.5.2-1Al, (Unit 1) 3.5.2-1B1, 3.5.2-132, and 3.5.2-1B3 (Unit 2), and 3.5.2-lcl, 3.5.2-1C2, 3.5.2-lC3 (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.
e.
Power level for Unit 1 shall not be greater than the power level cutoff unless one of the following requirements are met.
3.5-8 i
e I
,e
n (1) Quadrant tilt is less than or equal to 2.5 percent and the xenon reactivity is within 10 percent of the value for opera-tion at steady-state rated power.
(2)
Quadrant tilt is greater than 2.5 percent and the xenon reac-tivity is within 5 percent of the value for operation at steady-state rated power.
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 3.5.2-3A, 3.5.2-3B, ar.el 3.5.2-3C.
If the imbalance is not within the e.nvelope defined by Figure 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C, corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable 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 12/27/74
r Bas-a The power-imbalance envelope defined in Figures 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-4) such that the maximum clad temperature will not exceed the Final Acceptance Criteria.
Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary.
Operation in a situation that 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 The 25% + 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the apper 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
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 pocition limits.
The minimum available rod worth, consistant with the rod position limits, provides for achieving hot shut-down by reactor trip at any time, assuming the highest worth cottrol rod that is withdrawn remains in the full out position (1).
The roc position limits also ensure that inserted rod groups will not contain single rod worths greater than C.5% ak/k (Unit 1) or 0.65% A k/k (Units 2 and 3) at rated power.
These values have been shown to be safe by the safety analysis (2,3,4) 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-sequences than a 0.5% a k/k (Unit 1) or 0.65% a k/k (Units 2 and 3) ejected rod worth at rated power.
Control rod groups are withdrawn in sequence beginning with Group 1.
Groups j
1 5,6, and 7 are overlapped 25 percent.
The normal position at power is for j
Groups 6 and 7 to be partially inserted.
l l
- Actual operating limits depend on whether or not are used and their respective instrument and c alibrationincore or excore detectors errors. The method used to define the operating ilmits is defined in plant operating procedures.
3.5-10
!Q
~
The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using the definition of quadrant power tilt given in Technical Specifications, Section 1.6.
These limits in conjunction with the control rod position limits in Specification 3.5.2.5c ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densification.
The quadrant tilt and axial imbalance monitoring in Specifications 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.
Acceptance 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 and e to prevent excessive power peaking by transient xenon.
The xenon reactivity must be beyond the "undershoot" region and asymptotically approaching its equilibrium value at rated power.
REFERENCES 1 FSAR Section 3.2.2.1.2 1
2 FSAR Section 14.2.2.2 3FSAR SUPPLEMENT 9 4B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1)
BAW-1396 (UNIT 2)
BAW-1400 (UNIT 3) i 3.5-11
R00 POSITION LIMITS FOR 4 PUMP OPERATION APPLICABLE TO THE PERIOD FROM 5015 EFPC TO 25015 EFPO.
120 (173,102)
'(203.7,102)
(170,102) 100 RESTRICTED POWER REGION (160,92)
(204.2 LEVEL 92)
CUTOFF 80 OPERATION IN THIS (229.6,75)
REGION IS NOT (122,69)
ALLOUED Af 60
=
(300,52)
N PERMISSIBLE m
w 40 OPERATING REGION i
/
20 (0,5) 0 I
i I
I i
i e
i e
i i
0 50 100 150 200 250 300 ROD INDEX, 5 WITHORAWN O
2,5 5,0 7,5 10,0 0,
2,5 5,0 7,5 100 GROUP 5 GROUP 7 0
2,5 5,0 7,5 10,0 GROUP 6 Rod ind=1 is the percentagt sum of the withdrawal of Groups 5,6 and 7 UNIT 1 R0D POSITION LIMITS FOR 4 PUMP OPERATION 3.5-12 OCONEE NUCLEAR STATION Figure 3.5.2-1A.1
O
.)
ROD POSITICN LIMITS FCR 4 PLW OPERATION APPLICABLE TO THE PERIOD AFTER 25015 EFPD 120 (288. 1.
(257.5.102) 102) 100 (170.102)
)
RESTRICTED OPERATION IN THIS REGION REGION IS NOT 80 ALLOWED (234,73)
WITHDRAWAL LIMIT j60
=
~
(300,52)
E; PERMISSIBLE e
OPERATIf4G 40 REGION 5
.8 20 0
O 50 100 150 200 250 300 R00 INDEX, 5 WITHDRAWN O
2,5 5,0 7,5 10,0 0,
2,5 5,0 7,5 10,0 GROUP 5 GROUP 7 0
2,5 5,0 7,5 10,0 GROUP 6 l
Rod inder is the percentage sum of the withdrawal of Groups 5,6 and 7 UNIT 1 ROD POSITION LIMITS FOR 4 PUMP OPERATION 3.5-13 sumen OCONEE NUCLEAR STATION Figure 3.5.2-1A2
~,R J
ROD POSITION LIMITS FOR 4 PUMP OPERATION APPLICABLE DURING THE PERICO FROM 100 EFPO TO 250110 EFPD.
(177.4.i 2) 062.100 100 (222.3.102)
POWER LEVEL CUT 90 OPERkil0N IN THIS REGION OFF = 90",
[
IS NOT ALLOWED 80 N
5 (300,78) k 70 U
E 60 PERillSS.lBLE OPERATING E
A R
REGION 50 d=
J (108.8.48) as 5 40 p RESTRICTED REGION S*
30 20 -
(74.15) 10 -
0 O
50 100 150 200 250 300 90d index. ", Withdrawn 0
25 50 75 100 0
25 50 75 100 t
t i
i f
f f
f f
Group 5 Group 7 0
25 50 75 100 I
f f
?
I Group 6 Rod index is the percentage sum of the withdrawal of Groups 5,6 and 7 UNIT 2 R00 POSITION LIMITS FOR 4 PUMP OPERATION
' satrowt OCONEE NUCLEAR STATION 3.5-14 Figure 3.5.2-1B1
i
.R00 POSITION LIMITS FCR 4 PUMP CPERATION APPLICABLE DURING THE PERIOD FROM 250110 EFF0 TO 435t10 EFPD.
(177.4.102)
(1s2.so2)
(222.3.t02)
POWER LEVEL CUT 90 OPERATION IN THIS REGION g(172.3.87) 0FF = 90%
IS NOT ALLOWE0 80 D
(300.78) 70 U
~
E 60 PERillSSIBLE OPERATING REGION
- 50 (i s2.so) 40 TRICTED
$g
+
REGION Y
N 30 g
/
20 I"5)
(121.is) 10 (lis.s.o) 0 O
50 100 150 200 250 300 Rod Inder % Witnarawn 0
25 50 75 100 0
25 50 75 100 t
i 1
I i
i t
1 I
I Group 5 Group 7 0
25 50 75 100 1
1 1
t i
Group 6 Rod inden is the percentage sum of the withdrawal of Groups 5,6 nd 7 UNIT 2 R0D POSITION LIMITS FOR 4 PUMP OPERATION st OCONEE NUCLEAR STATION 3.5-14a Figure 3.5.2-182
G
^
..)
ROD POSITION LIMITS FOR 4 PLNP OPERATION APPLICABLE DURING THE PERICO AFTER 434t10 EFPD.
(162.102)
(270.102) 100 (173.8,90) 90 OPERATION IN THIS REGION (254.3,90)
[
l$ NOT ALLOWE0 80 POWER LEVEL CUT OFF
- 70 8
3
=
g 60 PERMISSIBLE OPERATING
- 50 h'f E
REGION (1s2.50)
J RESTRICTED 40 S
REGION
)
S i
~
30 g
j h
20 g
j (74.15)
(121.15) 10 (19.5.0) 0
'l e
i 0
50 100 150 200 250 300 Rod index. ", Witndrawn 0
25 50 75 100 0
25 50 75 100 t
t 1
1 I
I f
I f
f Group 5 Group 7 0
25 50 75 100 1
I f
f t
Group G Rod inder is the percentage sum of the withdrawal of Groups 5,6 and 7 UNIT 2 R00 POSITION LIMITS FOR 4 PUMP OPERATION OCONEE NUCLEAR STATION 3.5-15 Figure 3.5.2-1B3
'O
)
ROD POSITION LIMITS FOR 4 PUMP CPERATION APPLICABLE CURING THE PERICD FRCM 100 EFPD TO 250!10 EFPD.
(177.4.102)
(162.102)
(222.3.102) 100 -
l POWER LEVEL CUT 90 OPERATION IN THIS REGION
,[/
OFF = 90%
IS NOT ALLOWE0 80 N
~
(300,78) 8 70 E
=
$ 60 E
PERMISSIBLE OPERATING REGION 50 J
(108.8.48)
~
E A RESTRICTED a.
40 g
REGION N
30 20 (74.85) 10 0
O 50 100 ISD 200 250 300 Rod Index. $ Witndrann 0
25 50 7E 100 0
25 50 75 100 t
t t
I t
t 1
Group 5 Group 7 0
25 50 75 100 I
?
f I
t Group 6 Rod index is tne percentage sur.1 of the withdrawal of Groups 5,6 and 7 UNIT 3 R0D POSITION LIMITS FOR 4 PUMP OPERATION sent rowta OCONEE NUCLEAR STATION 3.5-16 Figure 3.5.2-1C1
O
'N j
J R00 POSITION LIMITS FOR 4 PlhF OPERATION APPLICABLE DURING THE PERICD FRCM 250!10 EFPD TO 435tlO EFPD.
(177.4.102)
I***'**' *'
100 POWER LEVEL CUT 90 OPERATION IN THIS REGION (172.3.871 0FF = 90;,
IS NOT ALLOWE0 80 p
(300.78) g 70
%g u
5 60
- g m
PERMISSIBLE OPERATING
- 50 h"[
REGION E
(162.50)
~
h@RESTRICTE0 E
REGION 8
a
~
30 20 (M.15)
(12i.15) 10 (sis.s.o) 0 i
O 50 100 150 200 250 300 Rod index. ", Withdrawn 0
25 50 75 100 0
25 50 75 100 i
f I
f f
f f
f I
I Group 5 Group 7 0
25 50 75 100 l
t f
i 1
Group 6 Rod index is the percentage sum of the withdrawal of Groups 5,6 and 7 UNIT 3 R0D POSITION LIMITS FOR 4 PUMP OPERATION isent OCONEE NUCLEAR STATION 3.5-16a Figure 3.5.2-1C2
R00 POSITION LIMITS FOR 4 PUMP OPERATION APPLICABLE DURING THE PERIOD AFTER 435110 EFPD.
(162.102)
(270,102) 100 -
(173.8.90) 90 -
OPERATION IN THIS RE010N (234.3.90)
IS NOT ALLOWE0 80 POWER LEVEL CUT OFF S
g 70 w
b g 60 4
PERWIS$1BLE OPERATING E
REGION
- 50 (is2.so) k
[ RESTRICTED %
~
REGION
[
f
~
~
8 30
/
20 (121.15) 10 (119.5.0)
O 0
50 100 150 200 250 300 Rod index, ". Withdrawn 0
25 50 75 100 0
25 50 75 100 e
t t
I i
t i
i f
Group 5 Group 7 0
25 50 75 100 I
f f
I I
Group 6 Rod index is tne percentage sum of the withdrawal of Groups 5,6 and 7 UNIT 3 ROD POSITION LIMITS FOR 4 PUMP OPERATION f OCONEE NUCLEAR STATIO 3.5-17 D
Figure 3.5.2-1C3 4
g
.s ROD POSITION LIMITS FOR 2 AND 3 PUMP OPERATION APPLICABLE TO THE PERIOD AFTER 50t5 EFPD 120 2 PUNP WITH0RAWAL LIMIT i
100 (125,102)
(150,102)
(230,102) (275,102) x OPERATION IN THIS S
REGION IS NOT y
ALLOWED (300,92) 1;;
80 RESTRICTED y
REGION 4
g 3 PUMP WITH0RAWAL LIMITS s
~
S y 60
[D us (300,64)
=
[+
5 E;
PERMISSIBLE OPERATING 1
w 40 REGION h
E 20 0
i i
i i
i i
i i
i i
0 50 100 150 200 250 300 R00 INDEX, 5 IITHORAWN 2,5 5,0 7,5 10,0 0,
2,5 5,0 7,5 10,0 GROUP 5 GROUP 7 0
2,5 5,0 7,5 10,0 GROUP 6 Rod inder is ins percentage sum of the withdrawal of Groups 5,6 and 7 UNIT 1 R00 POSITION LIMITS FOR 2 AND 3 PUMP OPERATIJN t
OCONEE NUCLEAR STATION 3.5-18 Figure 3.5.2-2A l
r RCD POSITICN LIM 7.TS FOR 2 AND 3 PtNP CPERATION (178.102) 100 -
90 OPERKTION IN THIS REGION IS NOT ALLOWE0 2 80 0
A E
E 70 O
8 M
kg 60 5
PERMISSIBLE OPERATING g
E REGION
^
E 50
( e s2.so) i m
sa RESTRICTE0 N40 REGION 8
s g
s%
~
g
[ 30 o
a d 20 g
(" ' 5 )
E 02i.is) 10 (lis.s.o) 0 O
50 100 150 200 250 300 Rod index. 5 Withdrawn 0
25 50 75 100 0
25 50 75 100 t
t t
t 1
I t
I I
Group 5 Group 7 0
25, 50, 75 100 i
i Group 6 Rod index is the percentage sum of the withdrawal of Groups 5,6 and 7 UNIT 2 R0D POSITION LIMITS FOR 2 AND 3 PUMP OPERATION OCONEE NUCLEAR STATION 3.5-19 Figure 3.5.2-2B
J R00 POSITICN LIMITS FCR 2 AND 3 PUMP CPERATION
( ns.ioz) 100 90 OPERATION IN THIS REGION S
l$ NOT ALLOWE0 h 80 E
t 5
u 70 a
Si E
60 PERMISSIBLE OPERATING g
g REGION
- 50 (is2.so) 5 40 RESTRICTED REGION 8
d
~
o 30 aE W 20 (n.15)
(121.15) 10 (lis.s.0)
Q t
f f
f I
f f
f f
f f
f 0
50 100 150 200 250 300 Rod Inder. 5 Withdrawn 0
25 50 75 100 0
25 50 75 100 i
t t
I f
f e
t t
i Group 5 Group 7 0
25 50 75 100 i
f f
f f
Group 6 Rod index is the percentage sum of the withdrawal of Grouos 5,6 and 7 UNIT 3 R0D POSITION LIMITS FOR 2 AND 3 PUMP OPERATION 3.5-20 OCONEE NUCLEAR STATION Figure 3.5.2-2C
.s
)
Power, % of 2568 MWt 110 102,-15.3
+14.1,102
~~
100 92,-22.1 90 80 70 69,-27.0 l
60
+28.1,52 50 40 1
1 i
i i
i
-30
-20
-10 0
10 20 30 Core imbalance, 5 QPERATIONAL POWER IMBALANCE ENVELOPE i
UNIT 1 i ut.Powta OCONEE NUCLEAR STATION Figure 3.5.2-3A l
3.d-21
c POWER, 2 of 2568 MWT
+7.1.102
-20.4,102,
-22.9,90
+14.8,90 80 _ _
60 --
+27.8,60
-35.0,60 40 --
20 --
1 I
I I
t g
g g
-40
-20 0
+20
+40 CORE IMBALANCE, %
I UNIT 2 OPERATIONAL POWER IMBALANCE ENVELOPE OCONEE NUCLEAR STATION 3.5-22 Figure 3.5.2-3B i
i
e.,
POWER, 8 of 2568 MWT
-20.4,102
+7.1,102 100 --
-22.9,90
+14.8,90 80 --
-35.0,60 60 --
+27.8,60 40 --
20 --
I I
i 1
f f
I g
-40
-20 0
+20
+40 CORE IMBALANCE, %
i UNIT 3 OPERATIONAL POWER IMBALANCE ENVELOPE OCONEE NUCLEAR STATION 4
3.5-23 Figure 3.5.2-3C
3 m
21 20 0 19 OCONEE I CYCLE 2
)
RESULTS
". 18
\\
5
\\
/N 17 A
.N 16
/
\\
h
[
GENERIC RESULTS 1
BAW 10103 2
15
=
s 14
._E
- j. 13 12 0
2 4
6 8
10 12 Axial Location of Peak Power From Bottom of Core, ft i
LOCA LIMITED MAXIMUM l
l ALLOWABLE LINEAR HEAT RATE
(
OCONEE NUCLEAR STATION 3.5-24 Figure 3.5.2-4 l
l l