ML15112B107
| ML15112B107 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/05/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duke Power Co |
| Shared Package | |
| ML15112B108 | List: |
| References | |
| DPR-38-A-121, DPR-47-A-121, DPR-55-A-118 NUDOCS 8305180045 | |
| Download: ML15112B107 (21) | |
Text
o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTO',
D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50- 269 OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 121 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 April 18, 1983, as supplemented April 22, 1983, 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 (iLthat 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.121 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
830518b045 830505 PDR ADOCK 05000269 P
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMISSION Jo n F. Stolz, Chi)
O erating Reactors Branch T4 ivision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 5, 1983
o
.UNITED STATES
.*1 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. 121 License No. DPR-47
- 1. The Nuclear Rdgulatory Commission (the Commission) has found that:
A. The application for amendment by Duke Power Company (the licensee) dated Apri'l 18, 1983, as supplemented April 22, 1983, 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 (iLthat 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:
3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 121 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 Jo n F. Stolz, Chief O erating Reactors Br ch #4 ivision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 5, 1983
UNITED STATES NUCLEAR 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. 118 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 April 18, 1983, as supplemented April 22, 1983, 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 (ij.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 issuan e 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 issuanc6 of this amendment is in accordance with 10 CFR Part 51 of the Commissioni'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. 118are 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 Joh F. Stolz, Chief Op rating Reactors Branch 4
- vision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 5, 1983
ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO. 121 TO DPR-38 AMENDMENT NO. 121 TO DPR-47 AMENDIIgNT NO. 118 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 4.1-3 4.1-3 4.1-4 4.1-4 4.1-5 4.1-5 4.1-6 4.1-6 4.1-8 4.1-8 4.4-6 4.4-6 4.4-9 4.4-9 4.4-10 4.4-1.0 4.4-11 4.4-11 4.4-13 4.4-13 4.6-1 4.6-1 4.18-1 4.18-1 Pages 4.4-5 and 4.4-14 are overleaf pages and are included to inaintain document completeness.
0D C+
Table 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 0
Channel Description Check Test Calibrate Remarks
.1. Protective Channel NA MO NA Coinicidence-Logic
- 2. Control Rod Drive NA MO NA Trip Breaker C3
- 3. Power Range Amplifier ES(l)
NA (1)
(1)Heat balance check each shift. Heat balance calibration whenever indi cated core thermal power exceeds neu power by more than 2 percent.
- 4. Power Range ES M0 MO(1)(2)
(1) Using incore instrumentation.
(2) Axial offset upper and lower chambers after each startup if not done pri vious week.
- 5.
Intermediate Range ES(1)
PS NA (1)
When in service.
- 6. Source Range ES(1)
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
CD
- 3 3
Table 4.1-1 (CONTINUED) 0 Channel Description Check Test Calibrate Remarks
- 12.
Pump-Flux Comparator ES Mo RF
- 13.
High Reactor Building DA-MO-RF-Pressure
- 14.
High Pressure Injection &
NA MO NA Includes Reactor Building Reactor Building Isolation Isolation of non-essential 00 Logic (Non-essential systems) systems
- 15.
High Pressure Injection Analog Channels:
- a. Reactor Coolant Pressure ES MO RF
- b. Reactor Building Pressure (4 psig)
ES MO RF
- 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 110 NA Reactor Building isolation Cooling and Isolation includes essential systems System Logic (Essential Systems)
- 19.
Reactor Building Emergency ES M0 RF Cooling and Isolation System Analog Channel Reactor Building Pressure (4 psig)
(D CL CD, Table 4.1-1 (CONTINUED) 0 Channel Description Check Test Calibrate Remarks
- 20.
Reactor Building NA MO NA System Logic
- 21. Reactor Building Spray NA MO RF System Analog Channel Reactor Building High Pressure co
- 22.
Pressurizer Temperature ES NA RF
- 23. Control Rod Absolute ES(l)
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 RE Injection Systems Flow Channels
(D (D
Table 4.1-1 (CONTINUED) 0 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 NO NA AN CO
- 32.
Concentrated Boric Acid)
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 o, recorder readout.
- 35.
Emergency Plant MO(1)
NA RF (1) Battery check.
Radiation Instruments
- 36.
EnvironmenNA RF (1) Check functioning.
36 Evioaetal Monitors N0(1
- 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
- 49.
Emergency Feedwater 1O NA RF Flow Indicators a
- 50.
PORV and Safety Valve MO NA RF Position Indicators
- 51.
RPS Anticipatory NA MO R--RF Reactor Trip System Loss
..of Turbine Emergency Trip System Pressure Switches
- 52.
RPS Anticipatory Reactor Trip System Loss of Hain Feedwater a) Control Oil Pressure NA
.0 RF Switches
- b) Discharge Pressure NA 10 RF Switches
- 53. Emergency Feedwater Initiation Circuits a) Control Oil Pressure NA MO RF Switches
- b) Discharge Pressure NA MO RF Switches ES -
Each Shift QU -
Quarterly DA -
Daily AN -
Annually WE Weekly PS -
Prior to startup, if not performed previous week M0 Monthly NA -
Not Applicable RE Refueling Outage This Technical Specification will become effective as follows:
Unit 1 -
at the first convenient outage prior to or at the end of Oconee 1 Cycle 8 Refueling 0itap Unit 2 - end of Oconee 2 Cycle 6 Refueling Outage Unit 3 - end of Oconee.3 Cycle 7 Refueling Outage During the interim period, these discharge pressure switches will be tested during cold shutdown not to exceed once per month.
- A one-time extension is granted for the instrument calibration such that it be performed during the l983 11nit 1 refueling outage, provided that such outage begins no later than July 16, 1983.
When contaiznent itegrity is established, the overall containment leak rate of 0.25 weight percent of containment air at 59 psig will assure that the limits of 10CFR100 will not be exceeded should the maximum hypothetical accident occur. In order to assure the integrity of the containment, periodic testing is performed at reduced pressure, 29.5 psig. The permissible leakage rate at this reduced pressure has been established from the initial integrated leak rate tests in conformance with 10CTR50-, Appendix J.
The containment air locks (i.e., Personnel Hatch and Emergency Hatch) are tested on a oore frequent basis than other penetrations. The air locks are utilized during periods of time when containment integrity is required as well as hen the reactor is shutdown. Proper verification of door seal integrity is required to ensure containment integrity. Because the door seals are recessed, damage from tools due to air lock entry is improbable; however, a leak test of the outer door seals has been shown to be an acceptable alternative to the full hatch test to ensure. air lock integrity.
REFERENC 5 (11)
- SAR, Sections 5 and 13.
Amendments Nos. 104, 104, & 101
TABLE 4.4-1 LIST OF PENETRATIONS WITH 10CFR50, APPENDIX J TEST REQUIREMENTS 0
PENETRATION TYPE A TEST NUMBER SYSTEM SYSTEM CONDITION LOCAL LEAK TEST REMARKS 1
Pressurizer liquid Note 1 Type C Note 2, 7b sample line (Unit 1 only) 2 OTSG A Note I Type C Note 7b Sample line 3
Component cooling Note 1 Type C Note 3, 7d inlet line 4
OTSG B Note 1 None required Note 7b drain line 5
RB normal Note 10 Type C Note 7a, 7b, 9 sump drain line 6
Letdown Note 1 Type C Note 2, 7b line 7
RC Pump seal Note 1 Type C Note 3, 7b, 9 return line 8
Loop A nozzle Not Vented None required Note 5, 7d warming line 9
RCS normal Not Vented None required Note 5 makeup line and HP injection
'A' loop 10 RC Pump Not Vented Type C Note 5, 7d, 9, 12 seal injection
TABLE 4.4-1 LIST OF PENETRATIONS WITH 10CFR50, APPENDIX J TEST REQUIREMENTS 0
PENETRATION TYPE A NUMBER SYSTEM SYSTEM CONDITION LOCAL LEAK TEST REMARKS 36 RB emergency Not Vented None required Note 5 37 sump recirculation line 38 Quench tank Note 1 Type C Note 2, 7d, 12 cooler inlet line 39 HP Nitrogen supply Note 1 None required Note 3 (manual valves)
(Unit 2, 3)
CFT Vent line Note 1 None required Note 3 (manual valves)
Only 40 RB emergency Note 1 None required sump drain line 41 Instrument-air Note 1 None required Note 3 (manual valves) supply & ILRT verification line.
42 SPARE Not in Use 43' OTSG A Note 1 None required Note 7b drain line 44 Component cooling Note 1 Type C Note 3, 7d to control rod drive inlet line 45 ILRT instrument Not Vented Type C Note 3, 7a line 46 Reactor head-wash.
Note 1 Type C Note 3 (manual valves) filtered water inlet
CD TABLE 4.4-1 LIST OF PENETRATIONS WITH 10CFR5O, APPENDIX J TEST REQUIREMENTS PENETRATION TYPE A TEST NUMBER SYSTEM SYSTEM CONDITION LOCAL LEAK TEST REMARKS 47 (Unit 1 Demineralized water Note 1 Type C Note 3, 7d only) supply to RC pump seal vents 48 Breathing air Note 1 None required Note 3 (manual valves) inlet 49 (Unit LP Nitrogen supply Note 1 None required Note 3 (manual valves) only) 50 OTSG A Emergency Not Vented None required Note 5 FDW line 51 ILRT Pressurization Note 1 None required Note 6a, 7a line 52 HP Injection to Not Vented None required Note 5
'B' loop 53 (All)
HP Nitrogen supply Note 1 None required Note 3 (manual valves) to 'A' core flood tank (Unit 2, 3) LP Nitrogen supply Note 2 None required Note 3 (manual valves) 54 Component Note 1 Type C Note 3, 7b, 9(8) cooling outlet line 55 Demineralized Note 1 Type C (Unit 1) Note 3, (manual valves), 12 water supply (Unit 2,3) Note 3, 9 (manual valves) 56 Spent fuel canal Note 1 None required Note 3 (manual valve) fill and drain 57 (Unit 1 DHR return Not Vented None required Note 4 only) line
CD CL (A
TABLE 4.4-1 LIST OF PENETRATIONS WITH 10CFR50, 0
APPENDIX J TEST REQUIREMENTS PENETRATION TYPE A TEST
-NUMBER--
SYSTEM SYSTEM CONDITION - -
LOCAL LEAK-TEST REMARKS__-_
ra 58 (All)
OTSG B Note 1 Type C Note 7b sample line (Unit 2, 3) Pressurizer sample Note 1 Type C Note 2, 7b line 59 CF tank Note 1 None required Note 2 sample line 60 RB sample Note 1 Type C Note 2, 7b, 9 line (outlet) 61 RB sample Note 1 Type C Note 3, 7b, 9 line (inlet) 62 (Units 2, DHR return Not vented None required Note 4 3 only) line Personnel Vented Type B Note 6b hatch Emergency Vented Type B Note 6b hatch Equipment Vented Type B Note 6c hatch Electrical Vented Type B Note 6a, 12 penetration
(D TABLE 4.4-1 rt NOTES (continued) 0
- c.
Isolation valves are required to operate intermittently under post accident conditions.
- d.
Check valves used for containment isolation.
NOTE 8 DELETED NOTE 9 Reverse direction test of inside containment isolation valve authorized. Leakage results are 00
-conservative.
NOTE 10 System is submerged during post-accident conditions and-performance of Type A test. System will be drained to the extent possible.
NOTE 11 Type B test performed on the blind flanges inside the Reactor Building. The tube drain valves and valves outside the containment are not tested.
NOTE 12 A one-time extension from the local leak test and corresponding exemption from Sections III.D.2 and III.D.3 of Appendix J to 10 CFR Part 50 is granted such that it be performed during the 1983 Unit 1 refueling outage, provided that such outage begins no later than July 16, 1983.
Si
4.4.2 Structural Integrity ADDlicabilict Applies to the structural integrity of the Reactor Building.
Objective To define the inservice surveillance program for the Reactor Building.
Soecification 4.4.2.1 Tendon Surveillance For the initial surveillance program, covering the first five years of operation, nine tendons shall be selected for periodic inspection for symptoms of material deterioration or force reduction. The surveillance tendons shall consist of three horizontal tendons, one n each of three 120q sectors of the containment; three vertical-tendons located at approximately 1200 apart; and three dome tendons located approximately 120* apart. The
.ollowing nine tendons have been selected as the surveillance tendons:
Dome 1D28 2D28 (Units 1
- 3) 2D29 (Unit 2) 3D28 Horizontal 13H9 5 1H9 53H10 Vertical 22V14 4..2.1.1 Lift-Off i.t-off readings shall be taken for all nine surveillance tendons.
W.4.2.1.2 ir Inspection and Testing One surveillance tendon of each directional group shall be relaxed and one wire from each relaxed tendon shall be removed as a samole and visually in sected for corrosion or oitting. Tensile tests shall also be performed on a minimum. o: :hree specimens taken from the ends and middle of each of the three
- izes.
The specimens shall be the maximum lngth cceptable for the test anoaratus :z be used and shall include areas representative or sig aiiicant corrosion or pitting.
After the wire removal, the tendons shall be retensioned to the stress level measured at the lift-off reading and then checked by a final lift-off reading.
Amendments Nos. 104, 104, & 101 4.4-14
4.6 EMERGENCY POWER PERIODIC"TESTING Applicability Applies to the periodic testing surveillance of the emergency power sources.
Objective To verify that <the emergency power sources and equipment will respond promptly and properly when required.
Specification 4.6.1 Monthly, a test of the Keowee Hydro units shall be performed to verify proper operation of these emergency power sources and associated equip-.
ment. This test shall assure that:
- a.
Each hydro unit can-be automatically started from the Unit 1 and 2 control room.
- b.
Each hydro unit can be synchronized through.the 230.kV overhead circuit to the startup transformers.
- c.
Each hydro unit can energize the 13.8 kV underground feeder.
- d.
The 4160 volt startup transformer main feeder bus breakers and standby bus breaker shall be exercised.
4.6.2
- a.
Annually, the Keowee Hydro units will be started using the emergency start circuits in each control room to verify that each hydro unit and associated equipment is available to carry load within 25 sec onds of a simulated requirement for engineered safety features.
- b.
Promptly following the above annual test, each hydro unit will be loaded to at least the combined load of the auxiliaries actuated by ESG signal in one unit and the auxiliaries of the other two units in hot shutdown by synchronizing the hydro unit to the offsite power system and assuming the load at the maximum practical rate.
4.6.3 Monthly, the Keowee Underground Feeder Breaker Interlock shall be verified to be operable.
4.6.4' *During each refueling outage' a simulated emergency transfer of the 4160 volt main feeder buses to the startup transformer (i.e., CT1, CT2 or CT3) and to the 4160 volt standby buses shall be made to verify proper operation.
4.6.5 Quarterly, the External Grid Trouble Protection System logic shall be tested to demonstrate its ability to provide an isolated power path between Keowee and Oconee.
4.6.6 Annually and prior to planned extended Keowee outages, it shall be demonstrated that a Lee Station combustion turbine can be started and
- A one-time extension is granted for the simulated emergency transfer of the 4160 volt main feeder buses such that it be performed during the 1983 Unit 1 refueling outage, provided that such outage begins no later than July 16, 1983.
Amendments Nos. 121
,121
& 118
.6-1
4.18 SNUBBERS Applicability Applies to hydraulic and mechanical snubbers used to protect the Reactor Coolant System and other safety-related systems.
Objective To verify that the required hydraulic and mechanical snubbers are operable.
Specification
.18.1 Each snubber associated with the.Reactor Coolant System and other safety-related systems, as specified in the appropriate Station Procedure shall be visually inspected. Visual inspections shall verify:
(i1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (31) in those locations where mechanical snubber movement can be manually induced, the snubbers shall be inspected as follows:
(a) At each refueling*, the inaccessible snubbers.shall be inspected near the beginning and the end of the outage.
(b) In the event of a severe dynamic event, snubbers, in that system which experienced the event shall be inspected during the refueling outage to assure that the snubbers have freedom of movement and are not frozen up.
The inspection shall consist of verifying freedom of motion using one of the following; (i) Manually induced snubber movement, (ii) evaluation of in place snubber piston setting; (iii) stroking the mechanical snubber through its full range of travel.
If one or more mechanical snubbers are found to be frozen up during this inspection, those snubbers shall be replaced (or overhauled) before returning to power. Re-inspection shall subsequently be
'performed according to the'schedule listed below.
Snubbers which appear inoperable as a 'result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other' snubbers that may be generically susceptible; and (2)' the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specification 4.18.4. However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be tested
- A one-time extension is granted for the inaccessible mechanical snubber inspection-such that it be performed during the 1983 Unit 1 refueling outage, provided that such outage begins no later than July 16, 1983.
4.18-1 Amendments Nos.
121,121
, & 118