ML19331E155
| ML19331E155 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/16/1975 |
| From: | Purple R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19331E154 | List: |
| References | |
| NUDOCS 8009050565 | |
| Download: ML19331E155 (21) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMIS$10N wasMINGTON. D. C. 20S55 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT *ID FACILITY OPERATING LICENSE Amendment No. 8 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 May 9, 1975, 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; 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 authori::ed 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 Cc,mmission's regulations; and 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.
2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR-38 is hereby amended to read as follows:
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"B.
Technical Specifications the Technical Specificatians contained in Appendices A and B, as revised, are F.ereby incorporated in the license.
The licensee shall opr< ate the facility in accordance with the Technical Specificatior.s, as revised by issued changes thereto through Change No. 18."
3.
This license amendment is effective as of the date of its issuance.
4 FOR THE NUCLEAR REGULATORY C05NISSION
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Robert A..Nrple, Chief Operating Reactors Branch #1 Division of Reactor Licensing
Attachment:
Change No. 18 to Technical Specifications Date of Issuance:
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W ASHINGTON. O. C. 20sS S DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT 70 FACILITY OPERATING LICENSE Amendment No. 8 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 May 9, 1975, 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; i
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of i
the Commission; C.
There is reasonable assurance (i) that the activities authori ed 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; and 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.
2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.8 of Facility License No. DPR-47 is hereby amended to read as follows:
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"B.
Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised, are hereby incorporated in the license.
He licensee shall operatc the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change No. 13."
3.
This license amendment is effective as of the date of in issuance.
FOR THE NUCLEAR REGULATORY C0bMISSION W. n _- _s Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing Attachsent:
Change No.13 to Technical Specifications Date of Issuance:
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e UNITEo STATES NUCLEAR REGULATORY COMMISSION wasMINGTON,0. C. 28555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 5 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 May 9, 1975, 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 authoriced 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; and j
D.
The issuance of this amendment will not be inimical to the common i
defense and security or to the health and safety of the public.
i 2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR 55 is hereby amended l
to read as follows:
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"B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised, are herevy incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change No.
S."
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C05NISSION Y
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Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing
Attachment:
Change No. S to Technical Specifications Date of Issuance:
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' ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 8 TO FACILITY LICEESE NO. DPR-38, CHANGE NO. 18 TO TECHNICAL SPECIFICATIONS; AMENDMENT NO. 8 TO FACILITY LICENSE NO. DPR-47, CHANGE NO. 13 TO TECHNICAL SPECIFICATIONS; AMENDMENT NO. 5 TO FACILITY LICENSE NO. DPR-55, CHANGE NO.
5 TO TECHNICAL SPECIFICATIONS; DOCKET NOS. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:
Remove Pages Insert Pages 3.1-8 3.1-8 3.1-9 3.1-9 3.5-7 3.5-7 3.5-8 3.5-8 3.5-10 3.5-10 3.5-11 3.5-11 3.5-14 3.5-14 3.5-14a 1
3.5-15 3.5-15 3.5-16 3.5-16 3.5-16a 3.5-17 3.5-17 3.5-19 3.5-19 3.5-20 3.5-20 e
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~ ' 3.1.3 Minimum Conditions fer Critienlity 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.
3.1.3.3 When the reactor coolant temperature is below the minimum temperature specified 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 subcritical by an amount equal to or greater
- than the calculated reactivity insertion due to depressurization.
3.1.3.4 The reactor shall be maintained subcritical by at least 1%4k/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, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach 8/8/5 to criticality.
The regulating rods shall then be positioned within their position limits defined by Specification 3.5.2.5 prior to l
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-i sequences are acceptable.
1 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 j
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 j
pressure of 900 psia is approximately 0.1%ak/k.
i 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 yover 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 NDIT of the primary ecolant system.
Heattp to this temperature will be accomplished by operating _the reactor coolant pumps.
If the shutdown margin required by Specificatien 3.5.2 is maintained, there is no possibility of an accidental criticality as a result of a decrease of coolant pressure.
3.1-8 JUL 161975
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.a The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1% suberitical vill assure that the reactor coolant system cannot bgep solid in the event of a red withdrawal accident of a start-up accident.53' 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 8/8/5 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, Answer 14.4.1 3.1-9
g.
If within one (1) hour of determination of an inoperable rod, it is not determined that a 1%Ak/k hot shutdown margin exists 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.
i.
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 tsactor 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 re positioned such that the rod that was declared inoperable 1. 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 8/8/5 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 Whenever the quadrant power tilt exceeds 4 percent, except for a.
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 1 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 full power for each I percent tilt below the power allowable for the reactor coolant pump combination as defined by Specification 2.3.
h (3) Except as provided in 3.5.2.4.b, the reactor shall be brou3 t 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 JUL 1 a 1975 y
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pump combination.
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.
d.
Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4.a(3) or 3.5.2.4.c above, subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the power range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each 1 per-cent tilt for the maximum tilt observed prior to shutdown.
Quadrant power tilt shall be monitored on a minimum frequency of a.
once every two hours during power operation above 15 percent of rated power.
3.5.2.5 Control Rod Positions 8/8/5 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 with-drawal limits
- are specified on Figures 3.5.2-1A1 (Unit 1),
3.5.2-1B1, 3.5.2-1B2 and 3.5.2-133 (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 achieve an acceptable control rod position. Acceptable control rod positions shall then be attained within two hours.
- d. Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-1A1 (Unit 1) [see additional operating restrictions for Unit 1]* 3.5.2-1B1, 3.5.2-132, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-lC2, 3.5.2-1C3 (Unit 3), unless the following requirements are met.
(1) The xenon reactivity shall be within 10 percent of the value fo.
operation at steady-state rated power.
1 (2) The. xenon reactivity shall be asymptotically approaching the value for operation at steady-state rated power.
3.5-8 JUL 16 575 1
e
- Base, The power-imbalance envelope defined in Figures 3.5.2-3A, 3.5.2-38, 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 applicatioa of:
a.
Nuclear uncertainty factors b.
Thermal calibration c.
Fuel densification effects d.
Hot rod manufacturing tolerance factors The 25% i 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper 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 1
rod worth. Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits. 'The minimum available rod worth, consistant with the rod position limits, provides for achieving hot shut-down by reactor trip at any time, a'ssuming the highest worth control rod 8/8/5 that is withdrawn remains in the full out position (1). The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.5% 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% ak/k (unit 1) or 0.65% ak/k (units 2 and 3) ejected rod worth at rated power.
Control rod groups are withdrawn in sequence beginning with Group 1.
Groups 5,6, and 7 are overlepped 25 percent. The normal position at power is for Groups 6 and 7 to be partially inserted.
- Actual operating limits depend on whether or not are used and their respective instrument and calibrationincore or excore detectors errors. The method used to define the operating limits is defined in plant operating procedures.
3.5-10 JUL 161975
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 limi,ts 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 provfie 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 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 2 FSAR Section.14.2.2.2 3FSAR SUPPLEMENT 9 4B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1) 8/8/5 BAW-1396 (UNIT 2)
BAW-1400 (UNIT 3) 3.5-11 JUL 161975 l
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1.
A00 INDEX 15 THE PEAC7.TAGE SUM OF THE WITHDRAWAL OF THE 2.
RESTRICTIONS ON WITHDAAWAL (HASHED AREAS) AAE N00lFIED AFTEA THE.CONTA0L A00 1NTEACHANGE (SEE FIGURE J.5.2-182) *
(l'25.102)
(134.102)
(242.102)
- 102 loo AESTRICTED AEGION 80 - 82.5 (182 3.82.5)
(262 7,82.5)
POWER LEVEL CUTCFF A
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50 10 0 15 0 200 250 300 Rod inden % withdrawal 0
25 50 75 10 0 i
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4 Grow 7 0
25 50 75 10 0 l
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Gene 6 CONTROL R0D GROUP WITHDRAVAL LIMITS Q
25 50 TS 00 FOR 4 PUMP OPERATION UNIT 2 I
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Geog 5 UNIT 2 3 5,.,'
's OCONEE NUCLEAR STATION 3.5-14 Figure 3 5.2-181 JUL 161975
1.
ROO INDEX 15 THE PERCENTAGE SUM OF THE WITH0RAWAL OF die CPERATING GROUPS.
2.
THE A00lfl0NAL RESTRICTIONS ON WITHORAWAL (HASHED AREAS) ARE IN EFFECT AFTER THE CONTROL A00 INTERCHANCE. THE RESTRICTIONS ON WITHORAWAL ARE FURTHER N00lFIED AFTER 435 FutL POWER DAvs CF OPERATION (SEE FIGURE 3 5 2-183)
(194,102)
(242,102)
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100 (172.4,82.5)
- 82.5
/
(300,82.5)
(182 3,82.5) 80 (262.7,82.5)
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POWER LEVEL h
RESTRICTED CUT 0FF 2
RECl0N h60 5
f (162,50) al g 40 n.
PERMIS$18LE OPERATING REGION 20 (121.15)
(119 5.0) 0 I
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50 -
10 0 ISO 200 250 300 Rod index, % Withdmwd 0
25 50 75 100 1
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Gme 7 0
25 50 75 10 0 L
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Gmp 6 CONTROL ROD GROUP WITHDRAWAL LIMITS 0
25 50 75 10 0
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I FOR 4 PUMP OPERATION UNIT 2 Gee 5 UNIT 2 en OCONEE NUCLEAR STATION Figure 3.5 2-182 JUL 161975
l.
ROO INDEX l$ THE PERCENTAGE $UM 0F THE VITHORAVAL OF THE OPERATING CROUPS.
2.
THE ADDITIONAL RESTRICTIONS ON WITH0RAWAL ARE IN EFFECT AFTER 4)$ FULL POWER DAYS OF OPERATION.
(291.4 (270.102) 102) 102 10 0
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(172.4.82 5)
- 2. 5 (30^.32.5) f (244.5,82.5)
~
80 RESTRICTED REGION PERMIS$18LE
~
OPERATING REGION S
(162.50)
Q=
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20 (121.15)
[ (189 5.0)
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50 10 0 15 0 200 250 300 Rod index, % Withdmwol 0
25 50 75 10 0 l
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1 Group 7 0
25 50 75 10 0 t
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Group 6 0
25 50 75 10 0 CONTROL LOD GROUP WITHDRAWAL :.lM!TS I
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I For. 4 PUMP OPERATION UNIT 2 Gmup 5 UNIT 2 OCONEE NUCLEAR STATION Figu.c 3.3.2-183,
3 5-15 JUL 161975
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1.
R00 INDEX is THE PERCENTAGE SUM OF THE WITHDRAWAL OF THE OPERATING CROUPS.
2.
RESTRICTICNS ON WITHDRAWAL (HASHED AREAS) ARE MODIFIED AFTER THE CONTROL ROD INTERCHANGE (SEE FIGURE 3.5.2-1C2) 4 (125,102)
(194,102)
(212.102) 100 RESTRICTED REC 10N 80 82.5
/
(182 9,82.5)
(262.7,82.5)
POWER LEVEL CUT 0FF a.
f 60 I
o PERMISSIBLE OPERATING 40 Y
RECl0N
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i 0
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50 10 0 15 0 200 250 300 Rod hden, % Withdrowol 0
25 50 75 10 0 l
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Gap 7 0
25 50 75 10 0 l
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Gop 6 CONTROL RCD GROUP VITH0RAWAL LIM f
f FOR'4 PUMP OPERATI0li UNIT 3 O
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Group 5 UNIT 3
'cuube\\ OCONEE NUCLEAR STATION 3.5-16 Figum 3.5.2-Icl JUL 161973
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o R00INDEI15THEPERCENTAGE$Uk0FTHEWITH0RAWALOFTHE 1,
OPERATING CROUPS.
THE A00lfl0NAL RESTRICTIONS ON WITH0RAWAL (HASHED AREAS) ARE IN 2.
EFFECT AFTER THE CONTROL R00 INTERCHANCE. THE RESTRICTIONS ON WITHORAWAL ARE FURTHER N00lFIED AFTER 435 FULL POWER DAYS OF OPERATION ($EE FIGURE 3 5 2-lC))
(194,102)
(242,102) i 10 0 (172,4,82.5)
(182.3,82.5)
(3c0.32.5) 82.5 (262,7,82.5)
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80 POWER LEVEL RESTRICTED
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CUTOFF REC 10N
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f 60
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0
'(162,50) 3 w) 40 ne PERMIS$l8LE g
OPERATING REGION 20 (128,15) l
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(119.5.0) j o
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50 10 0 15 0 200 250 300 Rod inder, */. Withdrowel O
25 50 75 10 0 I
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Group 0
25 50 75 10 0 1
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Grow 6 CONTROL ROD WITH0RAWAL LIMITS 0
25 50 75 30 0 FOR !+ PUMP OPERATION UNIT 3 I
Group 5 l
0 OCONEE NUCL UNIT 3 Figure 3.5.2-Ic2 3.5-16a JUL 161975
5 1.
R00 INDEX 15 THE PERCENTAGE SUM OF THE WITHORAWAL OF THE OPERATING GROUPS.
2.
THE ADDITIONAL RESTRICTIONS ON WITHORAWAL ARE IN EFFECT AFTER 435 FULL DAYS POWER OF QPERATION.
(291.4, (270,102) 102)
-- 102
~
(172.4,82 5)
- 82.5 (300.82.5)
J (244.5,82.5) 80 PCWER LEVEL CUTOFF RESTRICTED
{
~
me 60
's OPERATING REGION y
(162.50)
I b
40 d
20
~
(128,15)
{ (119 5,0)
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O 50 10 0 15 0 200 250 300 Rod Index, % Withdrowo:
0 25 50 75 10 0 1
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Gee 7
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Groie 6 0
25 50 75 10 0 I
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Group 5 CONTROL ROD GROUP tilTHORAWAL L1MlTS FOR 4 PUMP OPERATION UNIT 3 UNIT 3 Orc OCONEE NUCLEAR STATION Figure 3.5.2-Ic3 3 5-17 JUL 161975
l.
ROD INDEX 15 THE PERCENTAGE SUM OF THE WITHDRAWAL OF THE OPERATING GROUPS.
(177.4.102)
- 102 T
mo f
E so PERMISSIBLE y
RESTRICTED REGION OPERATING REGION r
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E 60 E
E (162,50) 2 m
j 40 d<
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20 (121.15)
(119.5.0)
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50 10 0 15 0 200 250 soo Rod,Index, % Withdrawai O
25 50 75 10 0 l
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f f
Group 7 0
25 50 75 10 0 I
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Group 6 0
25 50 75 10 0
.,NTROL ROD GROUP WITHDRAWAL LIMITS i
I 1
1 1
FOR 3 AND 2 PUMP OPERATION UNIT 2 Group 5 UNIT 2 me,cowin OCONEE NUCLEAR, STATION Figure 3 5 2-28 3 5-19 JUL 161975
1 p.
. s l.
RCD INDEX 15 THE PERCENTAGE SUM OF THE WITHDRAWAL OF THE OPERATING CRCUPS.
102 (177.4.102) 10 0 RESTRICTED PERMISSIBLE g
REGION CPERATING REGION g
80 a
60 u-5 (162,50) 2 3
40 N
o 20 (121.15)
(119.5,0)
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50 10 0 15 0 200 250 300 Rod hden, % Withdrawol O
25 50 75 10 0 1
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Grow 7 O
25 50 75 10 0 I
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Group 6 sONTROL R00 GROUP WITH0RAWAL LIMITS O
25 50 75 10 0 FOR 3 AND 2 PUMP OPERATION UNIT 3 Gene 5 0niront UNIT 3 OCONEE NUCLEAR STATION Figure 3.5 2-2c JUL 1 a 1975 3 5-20
.