ML15113A126
| ML15113A126 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/23/1984 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duke Power Co |
| Shared Package | |
| ML15113A127 | List: |
| References | |
| DPR-38-A-132, DPR-47-A-132, DPR-55-A-129 NUDOCS 8412060546 | |
| Download: ML15113A126 (28) | |
Text
"PSREGC, UNITED oUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 132 License No. DPR-38
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated September 11, 1984, as supplemented on October 22, 26 and November 1, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to.this license.
amendment, and paragraph 3.B of Facility Operating License No. DPR-38 is hereby amended to read as follows:
3.3 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.132 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
8412060546 841123 PDR ADOCK 05000269 P
9 0
-2
- 3. 'This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION o~A F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 23, 1984
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 132 License No. DPR-47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated September 11, 1984, as supplemented on October 22, 26 and November 1, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-47 is hereby amended to read as follows:
3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 132are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ohW F. Stolz, Chie 0
rating Reactors Branch #4 Division of Licensing
Attachment:
Chances to the Technical Specifications Date of Issuance: November 23, 1984
$ R 0
UNITED STATES CNUCLEAR REGULATORY COMMISSION
/.
WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 129 License No. DPR-55
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated September 11, 1984, as supplemented on October 22, 26 and November 1, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-55 is hereby amended to read as follows:
3.B Technical Specifications The Technical Specifications contained -in Appendices A and B, as revised through Amendment No.129 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 0
F. Stolz, Chief(
erating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 23, 1984
ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO.132 TO DPR-38 AMENDMENT NO.132 TO DPR-47 AMENDMENT NO.129 TO DPR-55 DOCKETS NOS.
50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment numbers and contain vertical lines indicating the area of change.
Remove Pages Insert Pages 2.1-1 2.1-1 2.1-2 2.1-2 2.1-3 2.1-3 2.1-7 2.1-7 2.3-2 2.3-2 2.3-3 2.3-3 2.3-8 2.3-8 2.3-11 2.3-11 3.5-15 (3 pages) 3.5-15 (3 pages) 3.5-18 (3 pages) 3.5-18 (3 pages) 3.5-21 (3 pages) 3.5-21 (3 pages) 3.5-24 (3 pages) 3.5-24 (3 pages) 3.5-27 (3 pages) 3.5-27 (1 page)
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 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-lA-Unit 1. If the actual pressure/temperature point is below 2.1-1B-Unit 2 2.1-IC-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 ex pressed 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 havq beqn generated using the BAW-2 & BWC critical heat flux (CHF) correlations l, The BAW-2 correlation applies to fuel batches 9B and 10C while the BWC correlation applies to batches 10B and 11.
The reactor coolant system flow rate utilized is 106.5 (prcent of the design flow (131.32 x 106 lbs/hr) based on four-pump operation.
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 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 DNB is not an observable parameter during reactor operation, the observable para meters of neutron power, reactor coolant flow, temperature, and pressure 2.1-1 Amendment Nos. 132 132 129
can be related to DNB through the use of the CHF correlations (1,3).
The BAW-2 and BWC correlations have 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 (BAW-2) or 1.18 (BWC). A DNBR of 1.30 (BAW-2) or 1.18 (BWC) corresponds to a 95 per cent probability at a 95 percent confidence 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-lA represents the conditions at which the minimum allowable DNBR is predicted to occur for the limiting combination of thermal power and reactor coolant pump configuration. The curve is based upon the design nuclear power peaking factors including potential effects of fuel densification.
The curves of Figure 2.1-2A are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and rod bowing:
- 1.
The combination of the radial peak, axial peak and position of the axial peak that yields no less than the CHE correlation limit.
- 2.
The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.5 kw/ft for 9B, and 10C, and 11 Batches of fuel and 17.6 kw/ft for the 10B gadolinia fuel Batch 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 of Figure 2.1-3A correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
A B&W topical report discussing the mechanisms and resulting effects of fuel rod bow has been approved by the NRC (').
The report concludes that the DNBR penalty due to rod bow is insignificant and unnecessary because the power production capability of the fuel decreases with irradiation. Therefore, no rod bow DNBR penalty needs to be considered for thermal-hydraulic analyses.
The maximum thermal power for three-pump operation is 88.8 percent due to a power level trip produced by the flux-flow ratio (74.7 percent flow x 1.08 =
80.67 percent power plus the maximum calibration and instrument error).
The maximum thermal power for other coolant pump conditions is produced in a similar manner.
Amendment Nos. 132 132
& 129 2.1-2
For each curve of Figure 2.1-3A a pressure-temperature point above and to the left of the curve would result in a DNBR greater than the CHF correlation limit or a local quality at the point of minimum DNBR less than the CHF correlation quality limit for that particular reactor coolant pump situation.
The curve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.
References (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.
(2) Oconee 1, Cycle 4 - Reload Report -
BAW-1447, March, 1977.
(3) Correlation of 15x15 Geometry Zircaloy Grid Rod Bundle CHF Data with the BWC Correlation, BAW-10143P, Part 2, Babcock & Wilcox, Lynchburg, Virginia, August 1981.
.(4)
Fuel Rod Bowing in Babcock & Wilcox Fuel Designs, BAW-10147P-A, Rev. 1, Babcock & Wilcox, May 1983.
Amendment Nos, 132 132
& 129 2.1-3
Thermal Power Level, %
120
(-32.4, 112)
ACPTBE10(31.5, 112)
M 1 0.704 PUMP M = -1.049 OPERATION 100 20
(-53.9, 93.3) 1(-32.4, 88.8) 90 88.8)(53.9, 88.5)
(31.5, 885398.)
JACCEPTABLE 4 & 3 PUMP 8
OPERATION 8
(-53.9, 70.1) 70
(-32.4 61.1)
(53.9, 65.3)
ACCEPTABLE 60 UNACCEPTABLE 4, 3 & 2 PUMP OPERATION 50
-40 1(53.9, 37.6) 30
,4 C";20 UNACCEPTABLE UNA Am En e T INo 12 1N OPERATION
-60 F
-20 0
20 40 60 Reactor Power Imbalance, %
CURVE RC FLOW (GPM) 1 374,880 2
280,035 3
183,690 CORE PROTECTION SAFETY LIMITS UNIT 1 Amendment Nos.
132,132
& 129 2.1-7 WEosOCONEE NUCLEAR STATION Figure 2.1-2A
During normal plant operation with all reactor coolant pumps operating, 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 and instrument 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 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 the minimum allowable value should a low flow condition exist due to any electrical mal function.
The power level trip setpoint 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 coolant flow rate decreases. The power level trip setpoint produced by the power-to-flow ratio provides overpower DNB pro tection for all modes of pump operation. For every flow rate there is a maximum 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 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 92.59% and power level is 100%.
- 2. Trip would occur when three reactor coolant pumps are operating if power is 80.67% and reactor flow rate is 74.7% or flow rate is 69.44% and power level is 75%.
- 3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.92% and reactor flow rate is 49.0% or flow rate is 45.37% and the power level is 49%.
The analyses to determine the flux-to-flow ratios account for calibration and instrument errors and the maximum variation in RC flow in such a manner as to ensure a conservative setpoint. A Monte-Carlo simulation technique is used to determine the combined effects of calibration and instrument uncertainties with the final string uncertainties used in the analyses corresponding to the 95/95 tolerance limits.
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 (power 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 2.3-2B -
Unit 2 2.3-2C -
Unit 3 Amendment Nos.
132,132
, &129 2.3-2
level trip and associated reactor power/reactor power-imbalance boundaries by 1.08% -
Unit 1 for 1% flow reduction.
1.07% - Unit 2 1.08% - Unit 3 Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below the minimum allowable value 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 setpoint is reached before the nuclear over power trip setpoint. 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 (2300 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 (11.14 Tout-47 06)
(1800) psig (11.14 Tou-4706) setpoints shown in Figure 2.3-lA have been established to maintain to DNB 2.3-1B 2.3-1C ratio greater than or equal to the minimum allowable value 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)
(11.14 Tout - 4746)
(11.14 Tout - 4746) out Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (618 0F) 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 setpoint of 620 0F.
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 in the absence of a low reactor coolant system pressure trip.
Amendment Nos.
132, 132, & 129 2.3-3
Thermal Power Level, %
120
(-17, 108)
(17, 108)
.M = 1 1
1.0 ACCEPTABLE 100 2 = -1.209 4 PUMP OPERATION
(-40, 85)
(-17, 80.67)1 17, 80.67 1
80 (40, 80.19)
ACCEPTABLE(
UNACCEPTABLE 4 & 3 PUMP UNACCEPTABLE OPERATION OPrRATION OPERATION 60
(-40, 57.67)
(-17, 52.92)1
_17, 52.92)
(40, 52.86)
ACCEPTABLE 4, 3 & 2 -- 40 PUMP
(-40, 29.92)
OPERATION I (40, 25.11)
-20 j
I I
I II I
I 1
-60
-40
-20 0
20 40 60 Reactor Power Imbalance, %
PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS Amendment Nos, 132 132
& 129 2.3-8 UNIT 1 upon GCONEE NUCLEAR STATION Figure 2.3-2A
0 O
Table 2.3-lA Unit 1 Rectr rtectiv Sy2mTi Setting Limits Four Reactor Three Reactor One Reactor Coolant Pumps Coolant Pump Operating Operating eah In r~
P Sget(Operating Power (Operating PowerEahLo er-10 Rated) 75% Rated)
(Operating Power Shutdown
(%4ate
)
aRat5.e10
.5Byass 9?,
- 1.
Nuclear Power lMax.
105.5 105.%
105.5Byas
(%105.5 5.0()
,r 2.
Nuclear Power Max. Based on2 Flow (2) and Imbalance 1.08 times flow 1.08 times flow 1.08 times flow
(% Rated)
(
minus reduction minus reduction 08 e d tBypassed due to imbalance due to imbalance minus reduction
- 3.
Nuclear Power Hax.
Based NA due to imbalance on Pump Honitors, (% Rated)
NA 55%
- 4.
High Reactor Coolant 2300 2300 System Pressure, psig, Max.
2300 1720 S.
Low Reactor Coolant 1800 System Pressure, paig, min.
1800 1800 1800
- 6.
Variable Low Reactor 4706)(1)
Coolant System Pressure (11.14 T out -
4706 u(1) outg,(11.14 u
T
)(11.14 Tou 4706)(l Bypassed Psig, Min.Ou
- 7.
Reactor Coolant Temp. F., Max.
618 High Reactor Building 4618 618 Pressure p sig, Max.
(1) Tout is in degrees Fahrenheit (0F).
(2) Reactor Coolant System Flow, %.
(3) Administratively controlled reduction set only during reactor shutdown.
(4) AutoluaLiCally set when other segments of tilt!
RI'S are Iypa ssed.
100 OPERATION (137, 102)
(270, 102)
(300,102)
NOT ALLOWED
'POWER LEVEL SHUTDOWN CUTOFF=100%FP MARGIN (264, 92)
S.
LIMIT 3
80 80 (200, 80)
OPERATION RESTRICT 60 cr.
(89, 50)
OPERATION 40 ACCEPTABLE S
0e 20 e.(45,
- 15) 0 I
I O
100 200 300 GR 5 III_
Rod Index, % Withdrawn 0
75 100 GR 6 1 1
0 25 75 100 GR 7 0 25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 TO 30.+ 10/0 EFPD Amendment Nos.
132 132
&129 3.5-15 UNIT 1 3.55row OCONEE NUCLEAR STATION Figure 3.5.2-1 (1 of 3)
100 OPERATION (177, 102)
(264, 102)
(300,102)
NOT ALLOWED POWER LEVEL SHUTDOWN.
(258,
- 92)
CUTOFF=100% FP Z
MARGIN 80 LIMIT OPERATION (226, 80)
RESTRICT 60 (133, 50) 4 0
4J 40 OPERATION ALLOWED 0
(70,
- 15) 0,8.4) 0 0 0100 200 300 GR 5 J
Rod Index, % Withdrawn 0
75 100 GR 6 I 1 0
25 75 100 GR 7I(
0 25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION
.FROM 30 + 10/0 TO 250 + 10 EFPO UNIT 1 roEAOCONEE NUCLEAR STATION Figure 3.5.2-1 (2 of 3)
Amendment Nos.
132,132
, &129
100 OP(216, 1 )
(262 102) 10 -
OPERATION SUpW(300,102)VE NOT ALLOWED SHUTDOWN POWER LEVEL MARGIN (250, 92)
CUTOFF=100/FP LIMIT 3:
80 (224, 80)
(150, 50) 4 0
40 OPERATION ACCEPTABLE 20 (83, 15) 0, 8.0) 0 0
100 200 3CO GR5 GR61Rod Index, % Withdrawn 0
175 100 GR 6
L__
0 25 75 100 GR 7 0.
25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 250 + 10 EFPO UNIT 1I OCONEE NUCLEAR STATION Figure 3.5.2-1 (3 of 3)
Amendment Nos. 132 132
& 129
100 o -
SHUTDOWN MARGIN 80 LIMIT (137
- 77)
(272, 77)
(300,77)
(264, 9 1
60
-OPERATION OPERATION NOT ALLOWED RESTRICTED (200, 60) 40 4J 40 a) 0 89, 38)
U Si OPERATION 20 AC CEPTABLE (45, 11.7) 0 0
0 0
100 200 300 GR 5 1
1 Rod Index, % Withdrawn 0
75 100 GR7I 0
25 75 100 GR 7 [
I 0
25 100 ROD POSITION LIMITS FOR THREE PUNP OPERATION FROM 0 TO 30 + 10/-0 EFPD UNIT 1 DUKEPows OCONEE NUCLEAR STATION Amendment Nos.
132,132
, & 129 3.5-18 Figure 3.5.2-4 (1 of 3)
Z2 OPERATION 100 NOT ALLOWED o
SHUTDOWN (177, 77)
(266, 77)
(300,77)
LIMIT (258, 69) 44 60 OPERATION RESTRICT (242, 60) 40 C4 133, 3s)
U C.)
20 OPERATION ACCEPTABLE (70, 11.7)
,6.8) 0~
0 I
0 100 200 300 GR 5 I 1
Rod Index, % Withdrawn 0
75 100 GR 6.1 I
0 25 75 100 GR 71 I
0 25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 30 + 10/-0 TO 250 + 10 EFPD UNIT 1 kuuon OCONEE NUCLEAR STATION Figure 3.5.2-4 (2 of 3)
Amendment Nos.
- 132, 132, & 129
OPERATION 100 RESTRICTED OPERATION 8
NOT ALLOWED (216,
- 77) (266,
- 77) (300,77) 8 80.
Tr u
SHUTDOWN (250,69)
S o60 MARGIN LIMIT 242, 60) 4-0
.4-., a 40 150, 38)
S 20 OPERATION ACCEPTABLE 0
(83, 11.7) 0 (0, 6.5) 0 100 200 ic0 GR 5 I
_Rod Index, % Withdrawn U
lb 10 GR 6 0
25 75 100 GR 7 I I
I 0
25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION AFTER 250 + 10 EFPD UNIT 1 OCONEE NUCLEAR STATION Figure 3.5.2-4 (3 of 3)
Amendment Nos.
132, 132, & 129
L.
100 0
OPERATION 80 NOT ALLOWED a
(137, 52)
(273,
- 52)
(300,52) o SHUTDOW OPRTN MARGIN OPRTO 40 -LMTRESTRICTED 264, 46)
(200, 40) 20
-89,
- 26) 0 OPERATION ACCEPTABLE 1
(45, 8.5) 0
(
- 9) '
1 0
100 200 300 GR 5 Rod Index, % Withdrawn 0
75 100 GR6 1 I
0 25 75 100 GR 7 1 1
0 25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION FROM 0 TO 30 + 10/-0 EFPD UNIT I 3-2unow OCONEE NUCLEAR STATION Amnden3.5-21
- Nos, Figure 3.5.2-7 Amendment Nos. 132 132 1(1 of 3)
100 0
OPERATION NOT ALLOWED 80 60 (177, 52)
(267, 52)
(300,52) 4-SHUTDOWN o
(258, 46) 40'.
MARGIN LIMIT 242, 40)
OPERATION (13, RESTRICTED 20 (235,
- 26)
OPERATION O
5.2)
(70, 8.5)
ACCEPTABLE 0
100 200 200 GR 5 Rod Index, % Withdrawn 0
75 100 GR 6
0 25 75 100 GR7 0
25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION FROM 30 + 10/-0 TO 250 + 10 EFPD UNIT I 08EPOWERJ OCONEE NUCLEAR STATION Figure 3.5.2-7 (2 of 3)
Amendment Nos, 132, 132
, & 129
3o 100 80 60 SHUTDOWN (216,52) (268,52)
(300,52) o MARGIN LIMIT 5
(250,46).
40 OPERATION (242,40)
OPERATION NOT ALLOWED (150,26 (235,26) 0 20 0
,5.0)
(83,8.5)
OPERATION ACCEPTABLE 0
100 200 GR 5 I I
Rod Index, % Withdrawn 0
75 100 GR 6 I I
0 25 75 100 GR 71 I
0 25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION AFTER 250 + 10 EFPD UNIT 1 OUXnowE)l OCONEE NUCLEAR STATION Figure 3.5.2-7
(.3 of 3)
Amendment Nos.
132 132, & 129
(-19,102)
(24,102) 100
(-20,92) 90 (25,92)
IL
(-27,80) o 8
3,0 OPERATION 70 ACCEPTABLE 60 ra 0:
50
'4 40 30 20 10
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power Imbalance, %
OPERATIONAL POWER IMBALANCE ENVELOPE FROM 0 TO 30 + 10/-0 EFPD UNIT I Amendment Nos.
132 132 129 3.5-24 OCONEE NUCLEAR STATION Figure 3.5.2-10 (1 of 3)
(-22,102)
(27,102) 100
(-23,92) 90 (28,92)
(-30,80) o80 (33,80)
OPERATION ACCEPTABLE E 70 60 50 0
4j) 30 o
-20 10 IL I
__II I
I I
I
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power Imbalance, %
OPERATIONAL POWER IMBALANCE ENVELOPE FROM 30 + 101-0 TO 250 + 10 EFPD UNIT 1
< uxtpowi OCONEE NUCLEAR STATION Figure 3.5.2-10 (2 of 3)
Amendment Nos.
132, 132
, & 129
-(-25,102)
(712 (27,*102) 100
(-26,92) 90 (28,92)
(-32,80) 80 (33,80)
OPERATION ACCEPTABLE 70 70 60 50 0
4j 40 30 20 10
-50
-40
-30
-20
-10 0
10 20 30 40 5U Axial Power Imbalance, %
OPERATIONAL POWER IMBALANCE ENVELOPE AFTER 250 + 10 EFPD E POWER OUCNEE NUCLEAR STATION Figure 3.5.2-10 (3 of 3)
Amendment Nos.
132 132 1 29
Figure 3.5.2-13 (Deleted)
[Note that no rod position limits exist for Unit 1 axial power shaping rods.]
Amendment Nos, 132, 132
, & 129 3.5-27