ML16138A654
| ML16138A654 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/26/1989 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A655 | List: |
| References | |
| NUDOCS 8902020320 | |
| Download: ML16138A654 (24) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 172 License No. DPR-38
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility) Facility Operating License No. DPR-38 filed by the Duke Power Company (the licensee) dated September 3, 1987, as supplemented on February.27, September 9, and September 20, 1988, 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 set forth in 10 CFR Chapter I; D. The issuance of this license 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 hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 3.B. of Facility Operating License No. DPR-38 is hereby amended to read as follows:
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-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 172, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -
I/II
Attachment:
Technical Specification Changes Date of Issuance: January 26, 1989 OFFICIAL RECORD COPY LA:PD.I-3 P
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0 GUNITED STATES I oNUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 172 License No. DPR-47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility) Facility Operating License No. DPR-47 filed by the Duke Power Company (the licensee) dated September 3, 1987, as supplemented on February 27, September 9, and September 20, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules arid regulations set forth in 10 CFR Chapter I; E. 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 set forth in 10 CFR Chapter I; D. The issuance of this license 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 hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 3.B. of Facility Operating License No. DPR-47 is hereby amended to read as follows:
-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 172, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/f David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -
I/II
Attachment:
Technical Specification Changes Date of Issuance: January 26, 1989 OFFICIAL RECORD COPY LA:P II-3 P P -3 OGC-O D:PD 116/8)8 1/4 /8
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 169 License No. DPR-55
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility) Facility Operating License No. DPR-55 filed by the Duke Power Company (the licensee) dated September 3, 1987, as supplemented on February 27, September 9, and September 20, 1988, 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 set forth in 10 CFR Chapter I; D. The issuance of this license 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 hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 3.B. of Facility Operating License No. DPR-55 is hereby amended to read as follows:
-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 169, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -
I/Il
Attachment:
Technical Specification Changes Date of Issuance: January 26, 1989 OFFICIAL RECORD COPY LA:PI I-3 3
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ATTACHMENT TO LICENSE AMENDMENT NO. 172 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 172 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 169 FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Page Insert Page ii i
v-a v-a vii vii viii viii 1.-6 1.-6 3.1-8 3.1-8 3.1-9 3.1-9 3.1-23 3.1-23 3.5-7 3.5-7 3.5-8 3.5-8 3.5-9 3.5-9 3.5-10 3.5-10 3.5-11 3.5-11 3.5-12 3.5-12
-2 Remove Page Insert Page 3.5-15 3.5-15 3.5-16 3.5-17 3.5-18 3.5-19 3.5-20 3.5-21 3.5-22 3.5-23 3.5-24 3.5-25 3.5-26 3.5-27 3.5-28 3.5-29 6-9.1
Section.
Page 1'.5.4 Instrument Channel Calibration 1-3 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 POWER DISTRIBUTION 1-4 1.6.1 Quandrant Power Tilt 1-4 1.6.2 Reactor Power Imbalance 1-4 1.7 CONTAINMENT INTEGRITY 1-4 1.8 RADIOLOGICAL EFFLUENT CONTROL 1-5 1.8.1 Source Check 1-5 1.8.2 Offsite Dose Calculation Manual (ODCM) 1-5 1.8.3 Process Control Program (PCP) 1-5 1.8.4 Solidification 1-5 1.8.5 Gaseous Radwaste Treatment System 1-5 1.8.6 Ventilation Exhaust Treatment System 1-5 1.8.7 Purge-Purging 1-5 1.8.8 Venting 1-6, 1.8.9 Member(s) of the Public 1-6 1.8.10 Unrestricted Area 1-6 1.9 CORE OPERATING LIMITS REPORT 1-6 2
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1-1 2.1 SAFETY LIMITS, REACTOR CORE 2.1-1 2.2 SAFETY LIMITS - REACTOR.COOLANT SYSTEM PRESSURE 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 2.3-1 3
LIMITING CONDITIONS FOR OPERATION 3.0-1 3.0 LIMITING CONDITION FOR OPERATION 3.0-1 3.1 REACTOR COOLANT SYSTEM 3.1-1 OCONEE -
UNITS 1, 2 and 3 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
Section Page 6.5 STATION OPERATING RECORDS 6.5-1 6.6 STATION REPORTING REQUIREMENTS 6.6-1 6.6.1 Routine Reports 6.6-1 6.6.2 Non-Routine Reports 6.6-4 6.6.3 Special Reports 6.6-5 6.7 ENVIRONMENTAL QUALIFICATION 6.7-1 6.8 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.8-1 6.9 CORE OPERATING LIMITS REPORT 6.9-1 OCONEE -
UNITS 1, 2, and 3 v-a Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
LIST OF FIGURES Figure Page 2.1-1 Core Protection Safety Limits - Units 1, 2, and 3 2.1-4 2.1-2 Core Protection Safety Limits -
Units 1, 2, and 3 2.1-5 2.3-1 Protective System Maximum Allowable Setpoints -
Units 1, 2.3-5 2, and 3 2.3-2 Protective System Maximum Allowable Setpoints - Units 1, 2.3-6 2, and 3 3.1.2-1A Reactor Coolant System Normal Operation Heatup 3.1-6 Limitations -
Unit 1 3.1.2-1B Reactor Coolant System Normal Operation Heatup 3.1-6a Limitations - Unit 2 3.1.2-1C Reactor Coolant System Normal Operation Heatup 3.1-6b Limitations - Unit 3 3.1.2-2A Reactor Coolant System Cooldown Normal Operation 3.1-7 Limitations -
Unit 1 3.1.2-2B Reactor Coolant System Cooldown Normal Operation 3.1-7a Limitations -
Unit 2 3.1.2-2C Reactor Coolant System Cooldown Normal Operation 3.1-7b Limitations - Unit 3 3.1.2-3A Reactor Coolant System Inservice Leak and Hydrostatic 3.1-7c Test Heatup and Cooldown Limitation -
Unit 1 3.1.2-3B Reactor Coolant System Inservice Leak and Hydrostatic 3.1-7d Test Heatup and Cooldown Limitation - Unit 2 3.1.2-3C Reactor Coolant System Inservice Leak and Hydrostatic 3.1-7e Test Heatup and Cooldown Limitation -
Unit 3 3.1-10-1 Limiting Pressure vs. Temperature Curve for 100 STD 3.1-22 cc/Liter H20 3.5.2-16 LOCA-Limited Maximum Allowable Linear Heat 3.5-30 3.5.4-1 Incore Instrumentation Specification Axial Imbalance 3.5-34 Indication 3.5.4-2 Incore Instrumentation Specification Radial Flux Tilt 3.5-35 Indication 3.5.4-3 Incore Instrumentation Specification 3.5-36 OCONEE -
Units 1, 2, and 3 vii Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
LIST OF FIGURES (CONT'D)
Figure Page 4.5.1-1 High Pressure Injection Pump Characteristics 4.5-4 4.5.1-2 Low Pressure Injection Pump Characteristics 4.5-5 4.5.2-1 Acceptance Curve for Reactor Building Spray Pumps 4.5-9 6.1-1 Station Organization Chart 6.1-7 6.1-2 Management Organization Chart OCONEE -
UNITS 1, 2, and 3 viii Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
1.8.8 VENTING Venting is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during Venting. Vent, used in system names, does not imply a venting process.
1.8.9 MEMBER(S) OF THE PUBLIC Member(s) Of The Public shall include all persons who are not occupationally associated with the plant.
This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.
1.8.10 UNRESTRICTED AREA An Unrestricted Area shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection.of indivi-_
duals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial institutional and/or recreational purposes.
1.9 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current reload cycle.
These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.
Plant operation within these core operating limits is addressed in individual specifications.
OCONEE -
UNITS 1, 2, and 3 1-6 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
3.1.3 Minimum Conditions for Criticality Specification 3.1.3.1 The reactor coolant temperature shall be above 525 0F 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 the criticality limit of 3.1.2-1A (Unit 1) 3.1.2-1B (Unit 2) 3.1.2-1C (Unit 3) 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%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, safety 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 the acceptable operating limits for regulating rod position provided in the CORE OPERATING LIMITS REPORT.
Bases At the beginning of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly posit vj at operating temperatures with the operating configuration of control rods. 1 Calculations show that above 525 0 F, the consequences 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 0F is prohibited except where necessary for low power physics tests.
h potential reactivity insertion due to the moderator pressure coefficient 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 ' and the small integrated dk/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 the limits of Specification 3.1.2.1 provides increased assurance that the proper rela OCONEE -
UNITS 1, 2, and 3 3.1-8 Anendment No. 172 (Unit 1)
Amendment No.
172 (Unit 2)
Amendment No. 169 (Unit 3)
tionship between primary coolant pressure and temperature 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 is no possibility of an accidental criticality as a result of a decrease of coolant pressure.
The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1% subcritical will assure that the reactor coolant system cannot becoe solid in the event of a rod withdrawal accident or a startup accident.
The requirement that the safety rod groups be fully withdrawn before criticality 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. The acceptable operating position limits for the regulating rods for the appropriate unit and cycle are determined in accordance with the approved methodology and provided in the CORE OPERATING LIMITS REPORT per Specification 6.9.
REFERENCES (1) FSAR, Section 4.3.2 (2) FSAR, Section 4.3.2.4 (3) FSAR, Section 15.3 OCONEE -
UNITS 1, 2, and 3 3.1-9 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
3.1.11 Shutdown Margin Specification The available shutdown margin during all system conditions except refueling shall be greater than 1% Ak/k with the highest worth control rod fully with drawn.
Bases A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently sub critical to preclude inadvertent criticality in the shutdown condition.
During power operation and startup the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion limits determined in accordance with the approved methodology and provided in the CORE OPERATING LIMITS REPORT per Specification 6.9.
During refueling conditions equivalent protection is provided in the require ments of Specification 3.8.4.
OCONEE -
UNITS 1, 2, and 3 3.1-23 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
- c.
If a control rod is declared inoperable by being immovable due to excessive friction or mechanical interference or known to be untrippable then:
- 1.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify that the shutdown margin requirement of Specification 3.5.2.1 is satisfied and,
- 2.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> place the reactor in the hot standby condition.
- d.
If a control rod is declared inoperable due to causes other than addressed in 3.5.2.2.c above then:
- 1.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the rod to operable status or,
- 2.
Continue power operation with the control rod declared inoperable and
- a.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify the shutdown margin requirement of Specification 3.5.2.1 with an additional allowance for the withdrawn worth of the inoperable rod and,
- b.
Either reactor thermal power shall be reduced to less than 60% of the allowable power for the reactor coolant pump combination within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow/imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of thermal power value allowable for the reactor coolant pump combination or,
- c.
Position the remaining rods in the affected group such that the inoperable rod is maintained within allowable group average limits of Specification 3.5.2.2.a and within acceptable operating rod position withdrawal/
insertion limits for regulating rod position provided in the CORE OPERATING LIMITS REPORT.
- e.
If more than one control rod is inoperable or misaligned, the reactor shall be shut down to the hot standby condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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 provided in the CORE OPERATING LIMITS REPORT.
3.5.2.4 Quadrant Power Tilt
- a.
Except for physics tests, the maximum positive quadrant power tilt shall not exceed the Steady State Limit of Table 3.5-1 during power operation above 15% full power.
OCONEE -
UNITS 1, 2, and 3 3.5-7 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
- b.
If the maximum positive quadrant power tilt exceeds the Steady State Limit but is less than or equal to the Transient Limit of Table 3.5-1, then:
- 1. Either the quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Steady State Limit or,
- 2.
The reactor thermal power shall be reduced below 100% full power by 2% thermal power for each 1% of quadrant power tilt in excess of the Steady State Limit, and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by 2% thermal power for each 1% tilt in excess of the Steady State Limit. If less than four reactor coolant pumps are in operation, the allowable thermal power for the reactor coolant pump combination shall be reduced by 2% for each 1% excess tilt.
- c.
Quadrant power tilt shall be reduced within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to within its Steady State Limit or,
- 1.
The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the re actor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.
- d.
If the quadrant power tilt exceeds the Transient Limit but is less than the Maximum Limit of Table 3.5-1 and if there is a.
simultaneous indication of a misaligned control rod then:
- 1.
Reactor thermal power shall be reduced within 30 minutes at least 2% for each 1% of the quadrant power tilt in excess of the Steady State Limit.
- 2.
Either quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Transient Limit or,
- 3.
The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the reactor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.
- e.
If the quadrant power tilt exceeds the Transient Limit but is less than the Maximum Limit of Table 3.5-1, due to causes other than simultaneous indication of a misaligned control rod then:
- 1. Reactor thermal power shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the reactor OCONEE -
UNITS 1, 2, and 3 3.5-8 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combina tion.
- f.
If the maximum positive quadrant power tilt exceeds the Maximum Limit of Table 3.5-1, the reactor shall be shut down within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the Nuclear Overpower Trip Setpoints allowable for the reactor coolant pump combination are re stricted by a reduction of 2% of thermal power for each 1% tilt for the maximum tilt observed prior to shutdown.
- g.
Quadrant power tilt shall be monitored on a minimum frequency of once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 15% full 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.
Except for physics tests, operating rod group overlap shall be 25% t 5% between two sequential groups.
If this limit is exceeded, corrective measures shall be taken immediately to achieve an acceptable overlap. Acceptable overlap shall be attained within two hours or the reactor shall be placed in a hot shutdown condition within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c.
Position limits are specified for regulating and axial power shaping control rods.
Except for physics tests or exercising control rods, the regulating control rod insertion/withdrawal limits shall be maintained within acceptable operating limits for regulating rod position provided in the CORE OPERATING~
LIMITS REPORT for the particular number of operating reactor coolant pumps (4, 3, 2).
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.
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 acceptable operating limits for reactor power imbalance provided in the CORE OPERATING LIMITS REPORT.
OCONEE -
UNITS 1, 2, and 3 3.5-9 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
If the imbalance is not within the acceptable envelope, 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 or his designated alternate.
OCONEE -
UNITS 1, 2, and 3 3.5-10 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
Bases Operation at power with an inoperable control rod is permitted within the limits provided. These limits assure that an acceptable power distribution is maintained and that the potential effects of rod misalignment on associated accident analyses are minimized. For a rod declared inoperable due to misalignment, the rod with the greatest misalignment shall be evaluated first.
Additionally, the position of the rod declared inoperable due to misalignment shall not be included in computing the average position of the group for determining the operability of rods with lesser misalignments. When a control rod is declared inoperable, boration may be initiated to achieve the existence of 1% Ak/k hot shutdown margin.
The power-imbalance envelope obtained in accordance with the approved methodology is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-16) 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 power spike factors (Units 1 and 2 only)
- d.
Hot rod manufacturing tolerance factors
- e.
Fuel rod bowing power spike factors The 25% t 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 rod)
Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument calibration errors. The method used to define the operating limits is defined in plant operating procedures.
OCONEE -
UNITS 1, 2,
and 3 3.5-11 Amendment No.
172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No.
169 (Unit 3)
The rod position limits obtained in accordance with the approved methodology 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 position limits.
The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod 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 0.65% Ak/k at rated power.
These values have been shown to be safe by the safety analysis (2,3,4, 5) of hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% Ak/k is allowed by the rod position 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 consequences than a 0.65% Ak/k ejected rod worth at rated power.
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 Group 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 7.50% for Unit 1, 7.50% for Unit 2, 7.50% for Unit 3. The limits in Specific ation 3.5.2.4 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 withdrawal 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 resulting from xenon transients and power maneuvers are inherently included in the limits determined in accordance with the approved methodology.
OCONEE -
UNITS 1, 2, and 3 3.5-12 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
Figures 3.5.2-1 Thru 3.5.2-15 (deleted)
OCONEE -
UNITS 1, 2, and 3 3.5-15 thru 3.5-29 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)
Amendment No. 169 (Unit 3)
6.9 CORE OPERATING LIMITS REPORT Specification 6.9.1 Core operating limits shall be established prior to each reload cycle, or prior to any remaining part of a reload cycle, for the following:
(1) Power Dependent Rod Insertion Limits for Specifications 3.1.3.5, 3.5.2.2.d.2.c, 3.5.2.3, and 3.5.2.5.c.
(2)
Power Imbalance Limits for Specification 3.5.2.6.
and shall be documented in the CORE OPERATING LIMITS REPORT.
6.9.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically:
(1) DPC-NE-1002A, Reload Design Methodology II, October 1985.
(2) NFS-1001A, Reload Design Methodology, April 1984.
6.9.3 The core operating limits shall be determined such that all applic able limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
6.9.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
OCONEE -
UNITS 1, 2, and 3 6.9-1 Amendment No. 172 (Unit 1)
Amendment No. 172 (Unit 2)1 Amendment No. 169 (Unit 3)