ML15112B044
| ML15112B044 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/25/1982 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duke Power Co |
| Shared Package | |
| ML15112B045 | List: |
| References | |
| DPR-38-A-109, DPR-47-A-109, DPR-55-A-106 NUDOCS 8204080038 | |
| Download: ML15112B044 (11) | |
Text
1,R REG, 4 NU 9 EAR 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.109 License No. DPR-38
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A. The applications for-amendment by Duke Power Company (the licensee) dated October 1, 1976 and July 8, 1977, as supplemented on May 26 and September 21, 1977, March 6 and April 27, 1978, May 15, May 30, and July 16, 1979-and December 4, 1980, comply 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 applications, 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.109 are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
8204080038 620325 PDR ADOCK 05000269 P
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION AoF F. Stolz, Chief Op rating Reactors Branch #4 vision of Licensing Attathment:
Changes to the Technical Specifications Date of Issuance: March 25, 1982
p REG,,
'e.
UNITED STATES NULEAR 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.109 License No. DPR-47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The applications for amendment by Duke Power Company (the licensee) dated October 1, 1976 and July 8, 1977, as supplemented on May 26 and September 21, 1977, March 6 and April 27, 1978, May 15, May 30, and July 16, 1979 and December 4, 1980, comply 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 applications, 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. 109 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 SJob F. Stolz, Chief Op ating Reactors Branch #4 vision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: March 25, 1982
UNITEDSTATES LEAR REGULATORY COMMISSI WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 106 License No. DPR-55
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A.. The applications for amendment by Duke Power Company (the licensee) dated October 1, 1976 and July 8, 1977, as supplemented on May 26 and September 21, 1977, March 6 and April 27, 1978, May 15, May 30, and July 16, 1979 and December 4, 1980, comply 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 applications, 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:
.3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.106 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 oh F. Stolz, Chief Op ating Reactors Branch #4 vision of Licensing Attathment:
Changes to the Technical Specifications Date of Issuance: March 25, 1982
ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO.
109TO DPR-38 AMENDMENT NO.
109TO DPR-47 AMENDMENT NO. 106TO 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 iii iii 4-1
- 4. 01 4.5-2 4.5-2 4.5-7 4.5-7 4.20-1
Section 3.4 SECONDARY SYSTEM DECAY HEAT REMOVAL 3.4-1 3.5 INSTRUMENTATION SYSTEMS 3.5-1 3.5.1 Operational Safety Instrumentation 3.5-1 3.5.2 Control Rod Group and Power Distribution Limits 3.5-6 3.5.3 Engineered Safety Features Protective System 3.5-28 Actuation petpoints 3.5.4 Incore Instrumentation 3.5-30 3.6 REACTOR BUILDING 3.6-1 3.7 AUXILIARY ELECTRICAL SYSTEMS 3.7-1 3.8 FUEL LOADING.AND REFUELING 3.8-1 3.9 RELEASE OF LIQUID RADIOACTIVE WASTE 3.9-1 3.10 RELEASE OF GASEOUS RADIOACTIVE WASTE 3.10-1 3.11 MAXIMUM POWER RESTRICTIONS 3.11-1 3.12 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST 3.12-1 3.13 SECONDARY SYSTEM ACTIVITY 3.13-1 3.14 SHOCK SUPPRESSORS (SNUBBERS) 3.14-1
.3.15 PENETRATION ROOM VENTILATION SYSTEMS 3.15-1 3.16 HYDROGEN PURGE SYSTEM 3.16-1 3.17 FIRE PROTECTION AND DETECTION SYSTEMS 3.17-1 4
SURVEILLANCE REQUIREMENTS 4.0-1 4.0 SURVEILLANCE STANDARDS 4.0-1 4.1 OPERATIONAL SAFETY REVIEW 4.1-1 4.2 STRUCTURAL INTEGRITY OF ASME CODE CLASS 1, 2
AND 3 COMPONENTS 4.2-1 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4.3-1 4.4 REACTOR BUILDING 4.4-1 Amendments Nos. 109, 109, & 106 iii
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 eekly 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 fora 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.
4.0.4 Inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be-performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by.10 CFR 50 Section 50.55a(g) (4) to the extent practicable within the limitations of design, geometry and materials of construction of the components.
Aeds 1 4.0-1 Amendments Nos. 109 0T9
& 106
Reactor Coolant System, verification shall be made that the check and iso lation valves in the core flooding tank discharge lines operate properly.
- b. The test will be considered satisfactory if control board indication of core flood tank level verifies that all valves have opened.
4.5.1.2 Component Tests 4.5.1.2.1 Pumps Quarterly, the high pressure and low pressure injection pumps shall.be started and operated to verify proper operation. Acceptable performance will be indicated if the pump starts, operates for 15 minutes, and the discharge pressure and flow are within +/- 10 percent of a point on the pump head curve. (Figures 4.5.1-1 and 4.5.1-2) 4.5.1.2.2 Valves -
Power Operated
- a. Valves LP-17, -18 shall only be tested every cold shutdown unless previously tested during the current quarter.
- b. During each refueling outage, low pressure injection pump discharge (engineered safety features) valves, low pressure injection discharge throttling valves, and low pressure injection discharge header crossover valves shall be cycled manually to verify the manual operability of these power-operated valves.
4.5.1.2.3 Check Valves Periodic individual leakage testing (a) of valves CF-12, CF-14, LP-47.and LP-48 shall be accomplished prior to power operation after every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after. maintenance, repair or replacement work is performed.
Whenever integrity of these valves cannot be demonstrated, the integrity of the remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of the other closed.valve located.
in the high pressure piping shall be recorded daily.
For the allowable leakge rates and limiting conditions for. operation, see Technical Specification 3.1.6.10.
Bases The Emergency Core Cooling Systems are the principle reactor safety featues in the event of loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.
To satisfy ALARA requirements, leakge may be measured indirectly (as from the performance of pressure ind.icators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capabble of demonstrating valve compliance with the leakage criteria.
Amendments Nos. 109,109
, &'106 4.5-2
(2) Verification of the engineered safety features function of the Low Pressure Service Water System which supplies coolant to the reactor building coolers shall be made to demonstrate operability of the coolers.
(b) The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly, the appropriate pump breakers have completed their travel, fans are running at half speed, LPSW flow through each cooler exceeds 1400 GPM and air flow through each fan exceeds 40,000 CFM.
4.5.2.2 Component Tests 4.5.2.2.1 Pumps Quarterly, the reactor building spray pumps shall be started and operated to verify proper operation. Acceptable performance will be indicated if the pump starts, operates for 15 minutes, and the measured discharge pressure and flow results in a point above the pump head curve.
(Figure 4.5.2-1).
Bases The Reactor Building Coolant System and Reactor Building Spray System are designed to remove heat in the containment atmosphere to control the.rate of depressurization in the containment. The peak transient pressure in the con tainment is not affected by the two heat removal systems.
Hence, the basis for the spray pump flow acceptance test is the flow rate required during re circulation (1,000 gpm).
The delivery capability of one reactor building spray pump at a time can be
-tested by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corres ponding pump.
Pump discharge pressure and flow indication demonstrate per formance.
With the pumps shut down and the borated water storage tank outlet closed, the reactor building spray injection valves can each be opened and closed by operator action. With the reactor building spray inlet valves closed, low pressure air or fog can be blown through the test connections of the reactor building spray nozzles to demonstrate that the flow paths are open.
The equipment, piping, valves, and instrumentation of the Reactor Building Cooling System are arranged so that they can be visually inspected.. The cooling units and associated piping are located outside the secondary concrete shield. Personnel can enter the Reactor Building during power operations to inspect and maintain this equipment.
The service water piping and valves out side the Reactor Building are inspectable at all times. Operational tests and inspections will be performed prior to initial startup.
Amendments Nos.109
,109
, & 106 4.5-7