ML20077E721
| ML20077E721 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/08/1994 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20077E723 | List: |
| References | |
| NUDOCS 9412130104 | |
| Download: ML20077E721 (9) | |
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UNITED STATES
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,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20 % 5 0001
'g.....,o DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE. INC.
DOCKET NO. 50-413 CATAWBA NUCLEAR STATION. UNIT 1 eBfF0 MENT TO FACILITY OPERATING LICENSE Amendment No.126 License No. NPF-35 4
1.
The Nuclear Regulatory Comission (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 August 25, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules anc regulations as set forth in 10 CFR Chapter I; 3.
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 3
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 Ond all applicable requirements have been satisfied.
9412130104 941208 PDR ADOCK 05000413 P
1
- 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 a. follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.126
, and the Environmental Protection Plan ccntained 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 f rom the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION d
He bert N. Berkow, Director Project Directorate 11-3 Division of Reactor Pro o-I/II m
Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: December 8, 1994
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'4 UNITED STATES 2
j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 2066Mc01
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DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY NO 1 q
PIEDMONT MUNICIPAL POWER AGENCY DOCVsET N0. 50-414 CATAWBA NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.120 4
License No. NPF-52 1.
The Nuclear Regulatory Commission (the Commission) has found that:
I 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 Hunicipal Power Agency No. I and Piedmont Municipal Power Agency (licensees), dated August 25, 1994, 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 er 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.
1 1
1 l '
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.
4 NPF-52 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 120, 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.
i 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 l
H ertN.Berab, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
December 8, 1994 4
4 3
ATTACHMENT TO LICENSE AMENDMENT NO. 126 FACILITY OPERATING LICENSE N0. NPF-35 QOCKET NO. 50-413 eBD TO LICENSE AMEN 0 MENT N0. 120 FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 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.
Femove Paaes Insert Paaes 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 B 3/4 7-2 B 3/4 7-2
- B 3/4 7-2a B 3/4 7-2a
- 0verflow page
4 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater i
pumps and associated flow paths shall be OPERABLE with:
a.
Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and e
j b.
One steam turbine-driven auxiliary feedwater pump capable of being p
powered from an OPERABLE steam supply system.
APPLICABIllTY: MODES 1, 2, and 3.
ACTION:
a.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l l
b.
With two auxiliary feedwater pumps inoperable, be in at least H0T STANDBY witin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
f c.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
j SURVEILLANCE REQUIREMENTS j
4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
l a.
At least once per 31 days by:
1)
Verifying that each non-automatic valve in the flow path that is 1
not locked, sealed, or otherwise secured in position, is in its correct position-2)
Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic centrol or when above 10% RATED THERMAL POWER; and 3)
Verifying that the isolation valves in the auxiliary feedwater pump suction lines are open and that power is removed from the valve operators on Valves CA-2, CA-7A, CA-98, and CA-IIA and that the respective circuit breakers are padlocked, i
i l'
CATAWBA - UNITS 1 & 2 3/4'7-4 A.Tendment No. 126 (Unit 1)
Amendment No. 120 (Unit 2)
PLANT SYSTEMS SVRVElllANCE_ REQUIREMENTS (Continued) l b.
At least once per 92 days on a STAGGERED TEST BASIS by:
1)
Verifying that each motor-driven pump develops a total dynamic head of greater than or equal to 3470 feet at a flow of greater than or equal to 400 gpm; and 2)
Verifying that the steam turbine-driven pump develops a total dynamic head of greater than or equal to 3550 feet at a flow of 4
greater than or equal to 400 gpm when the secondary steam supply pressure is greater than 600 psig** and the auxiliary feedwater pump turbine is operating at less than or equal to 3800 rpm.
1 The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
c.
At least once per 18 months during shutdown by:
1)
Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, 1
2)
Verifying that each motor-driven auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater
]-
Actuation test signal, j
3)
Verifying that the turbine-driven auxiliary feedwater pump steam supply valves open upon receipt of an Auxiliary Feedwater Actua-tion test signal, and 4)
Verifying that the valve in the suction line of each auxiliary feedwater pump from the Nuclear Service Water System automatically actuates to its full open position within less than or equal to 16 seconds
- on a Loss-of-Suction test signal.
4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTOOWN of greater than 30 days prior to entering MODE 2 by verifying normal flowpath to each steam generator.
- Includes a time delay of up to 6 seconds.
- This verification is not required to be performed until 24 houts after achieving greater than or equal to 600 psig in the secondary side of the steam generator.
i CATAWBA - UNITS 1 & 2 3/4 7-5 Amendment No.126 (Unit 1)
Amendment No.120 (Unit 2)
ELANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a feedwater line break accident with a worst case single active failure.
The Auxiliary Feedwater System is capable of delivering a total feedwater flow of at least 492 gpm at a pressure of 1210 psig to the entrance of at least two of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reac-tor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.
Verification of the steam turbine-driven pump total dynamic head should be deferred until suitable test conditions are established (i.e., greater than or equal to 600 psig in the secondary side of the steam generator). This deferral is required because until 600 psig is reached, there is insufficient steam pressure to perform the test.
3/4.7.1.3 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.
This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the assumptions used in the safety analyses.
3/4.7.1.4 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to:
(1) minimize the positive reac-tivity effects of the Reacter Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.
The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.
3/4.7.1.5 CONDENSATE STORAGE SYSTEM The OPERABILITY of the Condensato Storage System with the minimum water volume ensures that sufficient water is available to maintain the Reactor Coolant system at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> cooldown with steam discharge to the atmosp e concurrent with total loss-of-of fsight power. The contained water volume.imit includes an allow-ance for water not usuable because of tank discharge line location or other physical characteristics.
CATAWBA - UNITS 1 & 2 B 3/4 7-2 Amendment No.126 (Unit 1)
Amendment No.120 (Unit 2)
PLANT SYSTEMS BASES 3/4.7.1.6 STEAM GENERATOR POWER OPERATED RELIEF VALVES The Surveillance Requirement for the Main Steam power-operated relief valves (PORVs) nitrogen supplies ensures that the PORVs will be available to mitigate the consequences of a steam generator tube rupture accident concurrent with loss of offsite power. This assumes that the PORY on the ruptured steam generator is unavailable, and that the other two are used to cool the Reactor Coolant System inventory to less than the saturation temperature of the ruptured steam generator.
Concurrent with the requirement that a specific number of PORVs be OPERABLE is the requirement that the associated PORY block valves upstream be open or OPERABLE.
Should an associated PORV block valve be closed and inoper-able, the PORV downstream of that block valve should also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to OPERABLE status.
)
Additionally, if a PORV is inoperable and open, then the requirements of Technical Specification 3.6.3, Containment Isolation Valves, would apply in addition to Technical Specification 3.7.1.6.
214,1jL STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maxi-mum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RTNDT of 60*F and are sufficient to prevent brittle fracture.
l CATAW8A - UNITS 1 & 2 8 3/4 7-2a Amendment No.126 (Unit 1)
Amendment No. 120 (Unit 2)
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