ML15112A990
| ML15112A990 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/28/1981 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112A991 | List: |
| References | |
| NUDOCS 8102200814 | |
| Download: ML15112A990 (17) | |
Text
RE~(I 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. 91 License No.
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated December 29, 1980, 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 pro visions 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 satis fied.
- 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.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 91 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
81022Oo'1
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: January 28, 1981
0 UNITED STATES NUCLEAR REGULATOR Y-COMMISSION E
.WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50- 270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 91 License No. DPR47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated December 29, 1980, 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 pro visions 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 satis fied.
- 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. 91 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 Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: January 28, 1981
o0 UNITED STATES NUCLEAR REGULATOR Y-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. 88 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 December 29, 1980, 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 1; B. The facility will operate in conformity with the application, the pro visions 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 satis fied.
- 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.88 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 Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 28, 1981
0 0
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 91 TO DPR-38 AMENDMENT NO. 91 TO DPR-47 AMENDMENT NO.
88-TO'DPR-!55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Revise Appendix A as follows:
Remove Pages Insert Pages 4-1 4-1 41-3 4.1-3 4.1-4 4.1-4 4.1-5 4.1-5 4.1-6 4.1-6 4.1-7 4.1-7 4.1-8 4.1-8 4.1-9 4.1-9 4.4-4 4.4-4 4.8-1 4.8-1
4 SURVEILLANCE REQUIREMENTS 4.0 SURVEILLANCE STANDARDS Applicability Applies to surveillance requirements which relate to tests, calibrations and inspections necessary to assure that the quality of structures, systems and components is maintained and that operation is within the safety limits and limiting conditions for operation.
Objective To specify minimum acceptable surveillance requirements.
Specification 4.0.1 Surveillance of structures, systems, components and parameters shall be as specified in the various subsections to this Technical Specifi cation section, Section 4.0, except as permitted by Technical Specifi cations 4.0.2 and 4.0.3 below.
4.0.2 Minimum surveillance frequencies, unless specified otherwise, may be adjusted as follows to facilitate test scheduling:
Maximum Allowable Specified Frequency Interval Between Surveillances Five times per week 2 days Two times per week 5 days Weekly 10 days Bi-Weekly 20 days Monthly 45 days Bi-Monthly 90 days Quarterly 135 days Semiannually 270 days Annually
.18 months Refueling Outage-22 months, 15 days 4.0.3 If conditions exist such that surveillance of an item is not necessary to assure that operation is within the safety limits and limiting conditions for operation, surveillance need not be performed if such conditions continue for a length of time greater than the specified surveillance interval. Surveillance waived as a result of this specification shall be performed prior to returning to conditions for which the surveillance is necessary to assure that operation is within safety limits and limiting conditions for operation.
Amendments Nos.
91 91, & 88 4-1
(D Table 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 0
Channel Description Check Test Calibrate Remarks WP
- 1. Protective Channel NA MO NA Coincidence Logic
- 2. Control Rod Drive NA MO NA Trip Breaker
- 3. Power Range Amplifier ES(1)
NA (1)
(1) Heat balance check each shift. Heat balance calibration whenever indi cated core thermal power exceeds neutron power by more than 2 percent.
- 4. Power Range ES HO MO(l)(2)
(1) Using incore instrumentation.
(2) Axial offset upper and lower chambers after each startup if not done pre 41 vious week.
- 5.
Intermediate Range ES(1)
PS NA (1)
When in service.
- 6. Source Range ES(l)
PS NA (1) When in service.
- 7. Reactor Coolant ES MO RF Temperature
- 8. High Reactor Coolant ES MO RF Pressure
- 9. Low Reactor Coolant ES MO RF Pressure
- 10. Flux-Reactor Coolant ES.
MO RF Flow Comparator
- 11.
Reactor Coolant Pressure ES MO RF Temperature Comparator
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks 0
- 12.
Pump-Flux Comparator ES MO RF
- 13.
High Reactor Building DA MO RF Pressure
- 14.
High Pressure Injection NA MO NA Logic 00
- 15.
High Pressure Injection Analog Channels:
- a. Reactor Coolant Pressure ES MO HF
- b. Reactor Building Pressure (4 psig)
ES MO HF
- 16.
Low Pressure Injection NA MO NA Logic
- 17.
Low Pressure Injection Analog Channels:
- a. Reactor Coolant Pressure ES MO RF
- b. Reactor Building Pressure (4 psig)
ES MO RF
- 18.
Reactor Building Emergency NA MO NA Cooling and Isolation System Logic
- 19. Reactor Building Emergency ES MO RF Cooling and Isolation System Analog Channel Reactor Building Pressure (4 psig)
Table 4.1-1 (CONTINUED)
(U Channel Description Check Test Calibrate Remarks 0
- 20.
Reactor Building NA MO NA System Logic
- 21.
Reactor Building Spray NA MO RF System Analog Channel Reactor Building High Pressure
- 22.
Pressurizer Temperature ES NA RF
- 23. Control Rod Absolute ES(1)
NA RF(2)
(1) Check with Relative Position Indi Position cator.
(2) Calibrate rod misalignment channel.
- 24.
Control Rod Relative ES(1)
NA RF(2)
(1) Check with Absolute Position Indi Position cator.
- 25.
Core Flood Tanks::
(2) Calibrate rod misalignment channel.
- a. Pressure ES NA RF
- b. Level ES NA RF
- 26. Pressurizer Level ES NA RF
- 27.
Letdown Storage Tank DA NA RF Level
- 28. Radiation Monitoring WE(1)
MO QU (1) Check functioning of self-checking Systems feature on each detector.
- 29.
High and Low Pressure NA NA RF Injection Systems Flow Channels
Table 4.1-1 (CONTINUED) w Channel Description Check Test Calibrate Remarks
- 30. Borated Water Storage WE NA RF Tank Level Indicator
- 31.
Boric Acid Mix Tank:
- a. Level NA NA AN
- b. Temperature MO NA AN 00
- 32.
Concentrated Boric Acid Co Storage Tank:
- a. Level NA NA AN
- b. Temperature MO NA AN
- 33.
Containment Temperature NA NA RF
- 34.
Incore Neutron Detectors MO(1)
NA NA (1) Check functioning; including functioning of computer readout or recorder readout.
- 35.
Emergency Plant MO(1)
NA RF (1) Battery check.
Radiation Instruments
- 36.
Environmental Monitors MO(1)
NA RF (1) Check functioning.
- 37.
Reactor Manual Trip NA PS NA
- 38. Reactor Building Emergency NA NA RF Sump Level
- 39.
Steam Generator Water.Level WE NA RF
- 40.
Turbine Overspeed Trip NA NA RF
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks 0
- 41.
Engineered Safeguards NA RF NA Channel 1 lP Injection Manual Trip
- 42.
Engineered Safeguards NA RF NA Channel 2 HP Injection Manual Trip oo
- 43.
Engineered Safeguards NA RF NA 00 Channel 3 LP Injection Manual Trip
- 44.
Engineered Safeguards NA RE NA Channel 4 LP Injection Manual Trip
- 45.
Engineered Safeguards NA RE NA Channel 5 RB Isolation
& Cooling Manual Trip
- 46. Engineered Safeguards NA RF NA Channel 6 RB Isolation
& Cooling Manual Trip
- 47.
Engineered Safeguards NA RF NA Channel 7 Spray Manual Trip
- 48. Engineered Safeguards NA RF NA Channel 8 Spray Manual Trip
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 0.
ES - Each Shift QU - Quarterly DA - Daily AN - Annually WE - Weekly PS -
Prior to. startup, if not performed previous week MO - Monthly NA-Not Applicable RF -Refueling Outage 00 00
Table 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency
- 1. Control Rod Movement (1)
Movement of Each Rod Monthly
- 2. Pressurizer Safety Valves Setpoint Each Refueling
- 3. Main Steam Safety Valves Setpoint Each Refueling(4 )
- 4. Refueling System Interlocks Functional Prior to Refueling
- 5. Main Steam Stop Valves (1)
Movement of Each Stop Monthly Valve
- 6. Reactor Coolant System (2)
Evaluate Daily Leakage
- 7. Condenser Cooling Water Functional Each Refueling System Gravity Flow Test
- 8. High Pressure Service Functional Monthly Water Pumps and Power Supplies
- 9. Spent Fuel Cooling System Functional Prior to Refueling
- 10. High Pressure and Low (3)
Vent Pump Casings Monthly and Prior Pressure Injection System to Testing (1) Applicable only when the reactor is critical.
(2) Applicable only when the reactor coolant is above.200aF and at a steady state temperature and pressure.
(3)
Operating pumps excluded.
(4) Number of safety valves to be tested each refueling shall be in accordance with ASME CodesSection XI, Article IWV-3511, such that each valve is tested at least once every 5 years.
Amendments Nos. 91,
91,
& 38 4.1-9
4.4.1.3 Isolation Valve Functional Tests Quarterly, remotely-operated Reactor Building isolation valves shall be stroked to the position required to fulfill their safety function unless such operation is not practical during unit operation. The latter valves shall be tested during each refueling shutdown.
4.4.1.4 Refueling Outage Inspection A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be performed each refueling out age and prior to any integrated leak rate test, to uncover any evidence of deteri oration which may affect either the containment's structural integrity or leak tightness. The discovery of any significant deterioration shall be accompanied by corrective actions in accord with acceptable procedures, non-destructive tests and inspections, and local testing where practical, prior to the conduct of any integrated leak rate test. Results of the inspection shall be reported to the Commission within 90 days of completion.
4.4.1.5 Reactor Building Modifications Any major modification or replacement of components affecting the Reactor Building integrity shall be followed by either an integratedleak rate test or a local leak rate test, as appropriate,.and shall meet the acceptance criteria of 4.4.1.1.4 and 4.4.1.2.3, respectively.
Bases The Reactor Building is designed for an internal pressure of 59 psig and a steam-air mixture temperature of 2860 F. Prior to initial operation, the con tainment is strength tested at 115 percent of design pressure and leak rate tested at the design pressure; The containment is also leak tested prior to initial operation at approximately 50 percent of the design pressure. These tests verify that the leak rate 'from Reactor Building pressurization satisfies the relationships given in the specification.
The performance of a periodic integrated leak rate test during unit life provides a current assessment of potential leakage from the containment, in case of an accident that would pressurize the interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic test is to be performed without pre liminary leak detection surveys or leak repairs, and containment isolation valves are to be.closed in the normal manner. The test pressure of 29.5 psig for the periodic integrated leak rate test is sufficiently high to provide an accurate measurement of the leak rate and it duplicates the preoperational leak rate test at 29.5 psig. The specification provides a relationship for relating the measured leakage of air at 29.5 psig to the potential leakage at 59 psig. The frequency of the periodic integrated leak rate testis normally keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.
The specified frequency of periodic integrated leak rate tests is based on three major considerations. First is the low probability of leaks in the Amendments Nos.
91, 91, & 88 4.4-4
4.8 MAIN STEAM STOP VALVES Applicability Applies to the main steam stop valves.
Objective To verify the ability of the main steam stop valves to close upon signal and to verify the leak tightness of the main steam stop valves.
Specification 4.8.1 Using Channels A and B, the operation of each of the main steam stop valves shall be tested during each refueling outage to demon strate a closure time of one second or less in Channel A and a clo sure time of 15 seconds or less for Channel B.
4.8.2 The leak rate through the main steam stop valves shall not exceed 25 cubic feet per hour at a pressure of 59 psig and shall be tested during each refueling outage.
Bases The main steam stop valves limit the Reactor Coolant System cooldown rate and resultant reactivity insertion following a main steam line break accident.
Their ability to promptly close upon redundant signals will be verified during each refueling outage. Channel A solenoid valves are designed to close all four turbine stop valves in 240 milliseconds. The backup Channel B solenoid valves are designed to close the turbine stop valves in approximately 12 seconds.
Using the maximum 15 second stop valve closing time, the fouled steam generator inventories and the minimum tripped rod worth with the maximum stuck rod worth, an analysis similar to that presented in FSAR Section.14.1.2.9, (but con sidering. a blowdown of both steam generators) shows that the reactor will remain subcritical after reactor trip following a double-ended steam line break.
The main stop valves would become isolation valves in the unlikely event that there should be a rupture of a reactor coolant line concurrent with rupture of the steam generator feedwater header. The allowable leak rate of 25 cubic feet per hour is approximately 25 percent of total allowable containment leakage from all penetrations and isolation valves.
REFERENCES (1) FSAR Supplement 2, Page 2-7 (2) Technical Specification 4.4.1 Amendments Nos.
91, 91 &
88 4.8-1