ML16293A760
| ML16293A760 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/16/1993 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A763 | List: |
| References | |
| NUDOCS 9309230066 | |
| Download: ML16293A760 (14) | |
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pREG~j UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
201 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 February 23, 1993, as supplemented May 4, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations as 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 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 attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-38 is hereby amended to read as follows:
9309230066 930916 PDR ADOCK 05000269 P
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.201, 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 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
September 16, 1993
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 201 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 February 23, 1993, as supplemented May 4, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations as 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 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 attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-47 is hereby amended to read as follows:
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 201, 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 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: September 16, 1993
N.R REG(
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.198 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 February 23, 1993, as supplemented May 4, 1993 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations as 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 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 attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-55 is hereby amended to read as follows:
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 198, 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 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: September 16, 1993
ATTACHMENT TO LICENSE AMENDMENT NO. 201 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND.
TO LICENSE AMENDMENT NO. 201 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 198 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 Pages Insert Pages 3.6-2 3.6-2 3.6-2a 3.6-3 3.6-3 3.6-3a 4.4-2 4.4-2 4.4-5 / 4.4-6*
4.4-5**
4.4 4.4-13 Deletion Page
- overleaf page to be deleted
- no change
- 2.
For plant conditions when the Reactor Coolant System.
temperature is above 250*F and pressure is above 350 psig but the reactor is at or below hot shutdown, one Reactor Building Purge isolation valve on each penetration may be open for testing and/or maintenance per Specification 4.4.4.1 and 3.6.6.
- 3.
For plant conditions other than contained in Specification 3.6.3.b.1,.2 above,.with one or more Reactor Building purge valves open, the open valves shall be closed within one hour, or the plant shall be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Reactor Coolant System temperature below 250,F and pressure below.350 psig.
- c.
A containment isolation valve, other than a Reactor Building Purge isolation valve, may be inoperable provided either:
- 1.
The inoperable valve is restored to operable status within four hours.
- 2.
The affected penetration is isolated within four hours by the use of a deactivated automatic valve secured and locked in the isolated position.'
- 3.
The affected penetration is isolated within four hours by the use of a closed manual valve or blind flange.'
- 4.
The reactor is in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.6.4 The reactor building internal pressure shall not exceed 1.5 psig or a vacuum of five inches of Hg if the reactor is critical.
,3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual containment isolation valves which should be closed are closed and tagged.
OCONEE -
UNITS 1, 2 & 3 3.6-2 Amendment No. 201 (Unit 1)
Amendment No. 201 (Unit 2)
Amendment No. 198 (Unit 3)
' Penetration flow paths (except for the Reactor Building Purge flow path) may be unisolated intermittently.under administrative controls.
3.6.6 The combined leakage rate for all penetrations and valves shall be determined in accordance with Specification 4.4.1.2. If, based on the most recent surveillance testing results the combined leakage rate exceeds the specified value and containment integrity is required then,
- 1) corrective action of Specification 3.6.3.c is met, or
- 2) repairs shall be initiated immediately and conformance with specified value shall be demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or the reactor shall be in cold shutdown within an additional 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
OCONEE UNITS 1, 2 & 3 3.6-2a Amendment No. 201 (Unit 1)
Amendment No. 201 (Unit 2)
Amendment No. 198 (Unit 3)
BASES The Reactor Coolant System conditions of cold shutdown assure that no steam will be formed and hence no pressure buildup in the containment if the Reactor Coolant System ruptures.
The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence.
The reactor building is designed for an internal pressure of 59 psig and an external pressure 3.0 psi greater than the internal pressure. The design external pressure of 3.0 psi corresponds to a margin of 0.5 psi above the differential pressure that could be developed if the building is sealed with an internal temperature of 120oF with a barometric pressure of 29.0 inches of Hg and the building is subsequently cooled to an internal temperature of 80oF with a concurrent rise in barometric pressure to 31.0 inches of Hg. The weather conditions assumed here are conservative since an evaluation of National Weather Service records for this area indicates that from 1918 to 1970 the lowest barometric pressure recorded is 29.05 inches of Hg and the highest of 30.85 inches of Hg.
Operation with a personnel or emergency hatch inoperable does not impair containment integrity since either door meets the design specifications for structural integrity and leak rate. Momentary passage through the outer door is necessary should the inner door gasket be inoperative to install or remove auxiliary restraint beams on the inner door to allow testing of the hatch. The time limits imposed permit completion of maintenance action and the performance of a local leak rate test when required or the orderly shutdown and cooldown of the reactor. Timely corrective action for an inoperable containment isolation valve is also specified.
Penetration flow paths, except for the Reactor Building Purge flow path, may be opened on an intermittent basis under administrative controls. Per NRC Generic Letter 91-08, acceptable administrative control for opening a containment isolation valve includes (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close the valve in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valve and that this action will prevent the release of radioactivity outside the containment.
When containment integrity is established, the limits of 10CFR100 will not be exceeded should the maximum hypothetical accident occur.
OCONEE -
UNITS 1, 2 & 3 3.6-3 Amendment No. 201 (Unit 1)
Amendment No. 201 Unit 2)
Amendment No. 198 (Unit 3)
The Reactor Building purge system was designed to allow cleanup of the Reactor Building atmosphere. It is normally operated during a unit shutdown which will require entry into the Reactor Building. It is used to purge the Reactor.
Building with fresh air to reduce the contaminant levels within the Reactor Building atmosphere, thus reducing overall personnel exposure. At times, certain safety related functions necessitate entry into the Reactor Building prior to cold shutdown conditions. These include isolation of leaking primary coolant system valves and visual inspections following outages. Use of the purge system tends to minimize any personnel exposure while not significantly contributing to overall plant risk.
The Reactor Building Purge System is required to be isolated whenever the RCS temperature is above.250oF and pressure is above 350 psig. The maximum pressure limit of 350 psig is based on the Oconee Unit 1 NPSH curve for RC pump operation. This will give a reasonable operating margin for the pumps while operating the purge. The LCO allows one isolation valve to be open on each penetration at or below hot shutdown for testing or maintenance.
REFERENCES FSAR, Section 3.8 OCONEE UNITS 1, 2 & 3 3.6-3a Amendment No. 201 (Unit 1)
Amendment No. 201 (Unit 2)
Amendment No. 198 (Unit 3)
- b.
Has a duration sufficient to establish accurately the change in leakage between the Type A test and the supplemental test.
- c.
Requires the quantity of gas bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total leakage rate at Pa (59 psig) or Pt (29.5 psig).
4.4.1.1.5 Report of Test Results The results of periodic tests shall be the subject of a summary technical report which shall be submitted to the Commission within 90 days of completion of the test.
4.4.1.2 Local Leak Rate Testing 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for containment penetrations in accordance with the criteria specified in Appendix J of 10CFR50.
4.4.1.2.2 Frequency of Test Local leak rate tests shall be conducted with gas at a pressure of not less than 59 psig during each reactor shutdown for refueling or other convenient interval but in no case at intervals greater than 24 months.
4.4.1.2.3 Acceptance Criteria The combined leakage rate from all penetrations and isolation valves shall not exceed 0.125 weight percent of the postulated post-accident containment air mass per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 59 psig.
4.4.1.3 Reactor Building Modifications Any major modification or replacement of components affecting the Reactor Building integrity shall be followed by either an integrated leak rate test or a local leak rate test, as appropriate, and shall meet the acceptance criteria of 4.4.1.1.3 and 4.4.1.2.3, respectively.
4.4.1.4 Isolation Valve Functional Tests Inservice testing of ASME Code Class 1, 2 and 3 valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10CFR50, Section 50.55a(g)(4) to the extent practicable within the limitations of design, geometry and materials of construction of the components.
OCONEE UNITS 1, 2 & 3 4.4-2 Amendment No. 201 (Unit 1)
Amendment No. 201 (Unit 2)
Amendment No. 198 (Unit 3)
When containment integrity 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 10CFR50, Appendix J.
The containment air locks (i.e., Personnel Hatch and Emergency Hatch) are tested on a more frequent basis than other penetrations. The air locks are utilized during periods of time when containment integrity is required as well as when 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.
REFERENCE (1) FSAR, Sections 3.8.1.7.4, 6.2.4, and 14 Amendments Nos. 139, 139, 136 4.4-5
PAGES 4.4-6 THROUGH 4.4-13 DELETED Amendment No. 201 (Unit 1)
OCONEE UNITS 1, 2, and 3 Amendment No. 201 (Unit 2)
Amendment No. 198 (Unit 3)