ML20117G396
| ML20117G396 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/13/1996 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20117G399 | List: |
| References | |
| NUDOCS 9605210236 | |
| Download: ML20117G396 (11) | |
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q UNITED STATES s
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20066-0001 8
.... 4 DUKE POWER COMPANY HQRTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE. INC.
DOCKET NO. 50-413 CATAWBA NVCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE 4
Amendment No.144 License No. NPF-35 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-35 filed by the Duke Power Company, acting for itself, North Carolina Electric Membership Corporation and Saluda River Electric Cooperative, Inc. (licensees), dated January 12, 1996, as supplemented March 4, April 3 and April 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 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.
9605210236 960513 PDR ADOCK 05000413 P
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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 2.C.(2) of Facility Operating License No.
NPF-35 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A, as revised i
through Amendment No.144
, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
Duke Power Company shall operate the facility in accordance with the Technical i
Specifications and the Environmental Protection Plan.
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d 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 II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes 4
Date of Issuance: May 13,1996 i
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UNITED STATES p
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4001 DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY N0. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 3
CATAWBA NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.138 License No. NPF-52 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPF-52 filed by the Duke Power Company, acting for itself, North Carolina Municipal Power Agency No. I and Piedmont Municipal Power Agency (licensees), dated January 12, 1996, as supplemented March 4, April 3 and April 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 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 i
E.
The issuance of this amendment is in accordance with 10 CFR 4
Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
I
i
. 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 2.C.(2) of Facility Operating License No.
NPF-52 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.138
, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Duke Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
i FOR THE NUCLEAR REGULATORY COMMISSION b
He bert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: May 13,1996 1
M I
I ATTACHMENT TO LICENSE AMENDMENT N0.144 FACILITY OPERATING LICENSE NO. NPF-35 i
DOCKET NO. 50-413 AND TO LICENSE AMENDMENT NO.11R FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 d
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 Paaes Insert Paaes 4
1 2
3/4 6-2 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-12 3/4 6-12 3/4 6-13 3/4 6-13 B 3/4 6-1 B 3/4 6-1 l
B 3/4 6-2 B 3/4 6-2 1
l i
1
CQtLIAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:
An overall integrated leakage rate of less than or equal to L, 0.30%
a.
by weight of the 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,,14.68 psIg.
b.
A combined leakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,,
and A combined bypass leakage rate of less than 0.07 L, for all pene-c.
trations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P,.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With:
(a) the measured overall integrated containment leakage rate exceeding 0.75 !s, or (b) the measured combined leakage rate for all penetrations and bypas:; le3kage rate exceeding 0.07 L bined leakag or (c) the combined
, restore the overa rate to less than 0.75 L, and the com l
and valves subject to Type B and C tests to less than 0.60 L, and the combined bypass leakage rate to less than 0.07 L, prior to in, creasing the Reactor Coolant System temperature above 200'F.
SURVEILLANCE RE0VINEMENTS 4.6.1.2 Type A containment leakage rates shall be demonstrated as required by 10 CFR 50.54(o) and Appendix J of 10 CFR Part 50, Option B, as modified by approved exemptions, and in accordance with the guidelines of Regulatory Guide 1.163, September, 1995.
CATAWBA - UNITS 1 & 2 3/4 6-2 Amendment No.144 (Unit 1)
Amendment No.138 (Unit 2)
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CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) f a.
Deleted 1
b.
Deleted c.
The accuracy of each Type A test shall be verified by a supplemental test in accordance with Regulatory Guide 1.163, September, 1995.
d.
Type B and C tests shall be conducted, in accordance with 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option A, with gas at a pressure not less than P,14.68 psig, at intervals no greater than 24 months except for test $ involving:
1)
Air locks, 2)
Purge supply and exhaust isolation valves with resilient material seals, and 3)
Dual-ply bellows assemblies on containment penetrations between the containment building and the annulus.
CATAWBA - UNITS 1 & 2 3/4 6-3 Amendment No.
144 (Unit 1)
Amendment No.
138 (Unit 2)
CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.
APPLICABILITY: H0 DES 1, 2, 3, and 4.
ACTION:
With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperaturf above 200'F.
SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed prior to the T containment leakage rate test (reference Specification 4.6.1.2) ype A to verify no apparent changes in appearance of the surfaces or other abnormal degradation.
If the Type A test is performed at 10-year intervals, two additional inspections shall be performed at approximately equal intervals during shutdowns between Type A tests. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission within 15 days as a Special Report pursuant to Specification 6.9.2.
CATAWJA - UNITS 1 & 2 3/4 6-12 Amendment No.
144 Unit 1)
Amendment No.
138 Unit 2)
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CONTAINMENT SYSTEMS REACTOR BUILDING STRUCTURAL INTEGRITY l
LIMITING CONDITION FOR OPERATION 3.6.1.7 The structural integrity of the reactor building shall be maintained i
at a level consistent with the acceptance criteria in Specification 4.6.1.7.
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APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
4 i
With the structural integrity of the reactor building not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.
SURVEILLANCE REQUIREMENTS 4.6.1.7 The structural integrity of the reactor building shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the reactor building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation. If the Type A test is performed at 10-year intervals, two additional inspections shall be performed at approximately equal intervals during shutdowns between Type A tests. Any abnormal degradation of the reactor building detected during the above required inspections shall be reported to the Commission within 15 days as a Special Report pursuant to Specification 6.9.2.
CATAWBA - UNITS 1 & 2 3/4 6-13 Amendment No. 144 Unit 1)
Amendment No.138 Unit 2)
l 3/4.6 CONTAINMENT SYSTEMS BASES i
3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restric-tion, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.
1 1
3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rater ensure that the total
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containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P. As an added conservatism, the as-left overall integrated leakage rate is further limited to less than er equal to 0.75 L to account for possible degradation of the containment leaksge barriersbetweenleakagetests.
j The surveillance testing for measuring Type A leakage rates is consistent with the requirements of Appendix J of 10 CFR 50, Option B.
Type B and C tests are conducted in accordance with 10 CFR 50 Appendix J, Option A.
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3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.
Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that:
(1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 1.5 psig, and (2) the containment peak pressure does not exceed the design pressure of 15 psig during LOCA conditions.
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CATAWBA - UNITS 1 & 2 B 3/4 6-1 Amendment No. 144 (Unit 1)
Amendment No. 138 (Unit 2) j i
C 4
CONTAINMENT SYSTEMS
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BASES i
INTERNALPRESSURE(Continued) f The maximum peak pressure expected to be obtained from a LOCA event is
)
14.68 psig. The limit of 0.3 psig for initial positive containment pressure is consistent with the safety analyses.
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3/4.6.1.5 AIR TEMPERATURE i
The limitations on containment average air temperature ensure that:
(1) the containment air mass is limited to an initial mass sufficientl prevent exceeding the design pressure during LOCA conditions, and (2) y low to i
the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment. Measurements shall be made at all operating ventilation acit locations, whether by fixed or portable instruments, prior to determining ti.e average air temperature.
The containment pressure transient is sensitive to the initially contained air mass during a LOCA.
The contained air mass increases with decreasing temperature. The lower temperature limit of 100*F for the lower compartment-and 75'F (60'F when in MODE 2, 3 or 4) for the upper compartment will limit the peak pressure to 14.7 psig which is less than the containment design pressure of 15 psig. The upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the safety analyses.
3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 15 psig in the event of a LOCA.
A periodic visual inspection is sufficient to demonstrate this capability.
3/4.6.1.7 REACTOR BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment reactor building will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to provide:
(1) protection for the steel vessel from external missiles, (2) radiation shielding in the event of a LOCA, and (3) an annulus surrounding the steel vessel that can be maintained at a negative pressure during accident conditions. A visual inspection is sufficient to demonstrate this capability.
Amendment No.144 (Unit 1)
CATAWBA - UNITS 1 & 2 8 3/4 6-2 Amendment No.138 (Unit 2)