ML16293A777
| ML16293A777 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/30/1996 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML16138A819 | List: |
| References | |
| NUDOCS 9611010279 | |
| Download: ML16293A777 (12) | |
Text
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.
218 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 August 12, 1996, as supplemented by letter dated September 10, 1996, 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; 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.
9611010279 961030 PDR ADOCK 05000269 P
PDR__
-2
- 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:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 218, 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, and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/
H rbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: October 30, 1996
o0 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. 218 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-38 filed by the Duke Power Company (the licensee) dated August 12, 1996, as supplemented by letter dated September 10, 1996, 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; 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.
0 g
-2
- 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:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.218, 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, and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION He4 ert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
October 30, 1996
- ,kREG&j 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. 215 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-38 filed by the Duke Power Company (the licensee) dated August 12, 1996, as supplemented by letter dated September 10, 1996, 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; 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
- 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:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.215, 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, and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION H bert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
October 30, 1996
0 0
ATTACHMENT TO LICENSE AMENDMENT NO. 218 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO.218 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 215 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 INSERT 4.4-1 4.4-1 4.4-2 4.4-2 4.4-3 4.4-3*
4.4-4 4.4-4 4.4-5 4.4-5
- No change -
information only
0 0
4.4 REACTOR BUILDING 4.4.1 Containment Leakage Tests Applicability Applies to Containment leakage.
Objective To verify that leakage from the Reactor Building is maintained within allowable limits.
Specification 4.4.1.1 Integrated Leak Rate Tests The containment leakage rate shall be determined, as required by IOCFR50.54 (o) and I0CFR50, Appendix J, Option B, including any approved exemptions, using the guidelines of Regulatory Guide 1.163, dated September, 1995.
4.4.1.1.1 Acceptance Criteria The overall acceptance containment leakage rate is determined by the preoperational leakage rate test and shall not exceed L., 0.25 weight percent of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P.,.59 psig. Any leakage in excess of 50% of the total allowed containment leakage shall be demonstrated to be to the penetration room. Containment leakage prior to startup following a Type A test shall not exceed.75 L,.
4.4.1.2 Local Leak Rate Testing 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for the containment penetrations in accordance with the criteria specified in Appendix J of 10CFR5O, Option A.
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.
Oconee Units 1, 2, and 3 4.4-1 Amendment No. 2 jL_(Unit 1)
Amendment No. 218 (Unit 2)
Amendment No.
-r (Unit 3)
0 0
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 appropnate, and shall meet the acceptance criteria of 4.4.1.1.1 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, and 3 4.4-2 Amendment No. 218 (Unit 1)
Amendmbnt No. 2 l 8 (Unit 2)
Amendment No. 215 (Unit 3)
4.4.1.5 Containment Air Lock Testing 4.4.1.5.1 Scope of Testing The Personnel Air Lock and Emergency Air Lock shall be tested as required by the following:
4.4.1.5.2 Frequency of Test (a)
The Personnel Air Lock and Emergency Air Lock shall be tested semiannually at an internal pressure of not less than 59 psig.
(b)
Air locks opened during periods when containment integrity is not required shall be tested at the end of such periods by a full hatch leak test at not less than 59 psig. If the full hatch test has been performed within the previous 3 days, the leak test can be performed between the double seal of the outer door at not less than 59 psig.
(c)
When containment integrity is required, either a full hatch leak test or a leak test of the outer door double seal will be performed within 3 days of initial opening, and during periods of frequent use, at least once every 3 days. Each leak test will be performed at not less than 59 psig.
4.4.1.5.3 Acceptance Criteria The acceptance criteria for the air lock leakage test is as stated in Specification 4.4.1.2.3.
Oconee Units 1, 2, and 3 4.4-3 Amendment No. _j3.
(Unit 1)
Amendment No. 135 (Umt 2)
Amendment No. 132 (Unit 3)
Bases The Reactor Building is designed for an internal pressure of 59 psig and a steam-air mixture temperature of 286 0F. This corresponds to a post-accident containment atmosphere mass of 5.1277 x 105 ibm. Prior to initial operation, the containment was strength tested at 115 percent of design pressure and leak rate tested at the design pressure. The containment was also leak tested prior to initial operation at approximately 50 percent of the design pressure. These tests verified that the leak rate from Reactor Building pressurization satisfies the relationships given in the specification.
The NRC approved an amendment to 10CFR50, Appendix J, "Leak Rate Testing of Containment of Light-Water-Cooled Nuclear Power Plants", to implement a performance based option for leakage testing of containment.
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. In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic test is to be performed without preliminary leak detection surveys or leak repairs, and containment isolation valves are to be closed in the normal manner.
Leakage to the penetration room, which is permitted to be up to 50 percent of the total allowable containment leakage, is discharged through high efficiency particulate air (HEPA) and charcoal filters to the unit vent. The filters are conservatively said to be 90 percent efficient for iodine removal.
More frequent testing of various penetrations is specified as these locations are more susceptible to leakage than the Reactor Building liner due to the mechanical closure involved. Testing of these penetrations is performed with air or nitrogen. The basis for specifying a maximum leak rate of 0.125 percent from penetrations and isolation valves is that one-half of the actual integrated leak rate is expected from those sources. Valve operability tests are specified to assure proper closure or opening of the Reactor Building isolation valves to provide for isolation of functioning of Engineered Safety Features systems.
Oconee Units 1, 2, and 3 4.4-4 Amendment No. 218 (Unit 1)
Amendment No. 2j8(Unit 2)
Amendment No. 215 (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.
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.
REFERENCES (1) FSAR, Sections 3.8.1.7.4, 6.2.4, and 14.
(2) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 1 OCFR Part 50, Appendix J", Revision 0; July 26, 1996 (3) Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program";
September 1995.
(4) NUREG 1493, "Performance-Based Containment Leak-Test Program", Revision 0, September 1, 1995.
Oconee Units 1, 2, and 3 4.4-5 Amendment No. 218 (Unit 1)
Amendment No. 218 (Unit 2)
Amendment No. 215j(Unit3)