ML20078D528
| ML20078D528 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/27/1994 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078D531 | List: |
| References | |
| GL-90-06, GL-90-6, NUDOCS 9411040257 | |
| Download: ML20078D528 (11) | |
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o UNITED STATES 5
NUCLEAR REGULATORY COMMISSION y..... j[
t WASHINGTON, D.C. 205SW1 DUKE POWER COMPANY l
QQCKET NO. 50-369 McGUIRE NUCLEAR STATION.' UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE
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Amendment No.150 License No. NPF-9 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Facility Operating License No. NPF-9 filed by the Duke Power Company (licensee) dated November 21, 1991, 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.
9411040257 941027 PDR ADOCK 05000369 P
. 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-9 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 150, are hereby incorporated into this license.
The licensee shall operate the faci'iity in accordance with the Technical Specifications and the Environmental Protection P1an.
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 Herbert N. Berkow, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nucle.ar Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
October 27, 1994
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0 UNITED STATES it NUCLEAR REGULATORY COMMISSION j/
WASHINGTON, D.C. 20555-0001
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DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NVCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 132 License No. NPF-17 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated November 21, 1991, 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.
j
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-17 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 132, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Technical Specifications and.the Environmental Protection P1an.
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 VN Herbert N. Berkow, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
October 27, 1994
ATTACHMENT TO LICENSE AMENDMENT NO. 150 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET N0. 50-369 AND TO LICENSE AMENDMENT NO. 132 FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 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 3/4 4-10 3/4 4-10 3/4 4-10a 3/4 4-37 3/4 4-37 3/4 4-38 3/4 4-38 B 3/4 4-3 8 3/4 4-3 8 3/4 4-3a B 3/4 4-3a*
- no change, overflow page t
REACTOR C00 ANT SYSTE'j A
3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION a.
With one or more PORV(s) inoperable because of excessive leakage, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and maintain power to the block valve (s); otherwise, 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 />.
b.
With one PORV inoperable due to causes other than excessive leakage, within I hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, 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 H0T SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With two PORVs inoperable due to causes other than excessive leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORVs to OPERABLE status or close the associated block valves and remove power from the block valves; otherwise, 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 />.
If the block valves have been closed and power has been removed, restore at least once PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in H0T 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 />.
d.
With three PORVs inoperable due to causes other than excessive leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLE status or close the associated block valves and remove power from the block valves and be in H0T 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 H0T SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
e.
With one block valve inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valve to OPERABLE status or place its associated PORV switch in the "close" position and remove power from its associated solenoid valve (do not enter action statement b for the resulting inoperable PORV);
otherwise, 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 />.
McGUIRE - UNITS 1 and 2 3/4 4-10 Amendment No.150 (Unit 1)
Amendment No.132 (Unit 2)
REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES (continued)
LIMITING CONDITION FOR OPERATION f.
With two block valves inoperable, within I hour restore the block valves to OPERABLE status or place their associated PORV switches in the "close" position (do not enter action statement c for the resulting inoperable PORVs); otherwise, be in at least H0T 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 />.
If the PORV switches have been placed in the "close" position, restore at least one block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in 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 H0T SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
9 With three block valves inoperable, within I hour restore the block valves to OPERABLE status or place their associated PORV switches in the "close" position (do not enter action statement d for the resulting inoperable PORVs).
Restore at least one block valve to OPERABLE status within the next hour; otherwise, be in H0T 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 />.
h.
The provisions of Specification 3.0.4 are no applicable.
SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:
a.
Performance of a CHANNEL CALIBRATION, and b.
Operating the valve through one complete cycle of full travel during MODE 3 or MODE 4 when the temperature of all RCS cold legs is greater than 300F with the block valve closed.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION a.,
b.,
c.,
or d. in Specification 3.4.4.
4.4.4.3 The emergency power supply for the PORVs shall be demonstrated OPERABLE at least once per 18 months by:
a.
Manually transferring motive power from the normal (air) supply to the emergency (nitrogen) supply.
b.
Isolating and venting the normal (air) supply, and c.
Operating the valves through a complete cycle of full travel.
MCGUIRE - UNITS 1 AND 2 3/4 4-10a Amendment No.150 (Unit 1) l 132 (Unit 2)
Amendment i
i
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:
a.
Two power-operated relief valves (PORVs) with a lift setting of less than or equal to 400 psig, or b.
The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 4.5 square inches.
APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 300 F, MODE 5, and MODE 6 when the head is on the reactor vessel.
ACTION:
a.
With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or complete depressurization and venting of the RCS through at least a 4.5 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b.
With one PORV inoperable in MODE 5, suspend all operations that could lead to water-solid RCS conditions.
Restore the inoperable PORV 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 complete depressurization and venting of the RCS through at least a 4.5 square inch vent (s) within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
With one PORV inoperable in MODE 6, restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or complete depressurization and venting of the RCS through at lest a 4.5 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, d.
With both PORVs inoperable, complete depressurization and venting of the RCS through at lest a 4.5 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
e.
In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.
The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient, and any corrective action necessary to prevent recurrence.
f.
The provisions of Specification 3.0.4 are not applicable.
MCGUIRE - UNITS 1 AND 2 3/4 4-37 Amendment No.150 (Unit 1)
Amendment No.132 (Unit 2)
REACTOR C0OLANT SYSTEM SURVEILLANCE RE0VIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
a.
Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days; b.
Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and c.
Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.
4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- when the vent (s) is being used for overpressure protection.
l i
- Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
MCGUIRE - UNITS 1 AND 2 3/4 4-38 Amendment No. 150 (Unit 1)
Amendment No. 132 (Unit 2)
REACTOR C0OLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.
Each P00V has ; remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:
1)
Manual control of PORVs to control RCS pressure. This is a function that is used for the steam generator tube rupture accident coincident with a loss of all offsite power and for plant shutdown.
- 2) Maintaining the integrity of the reactor coolant pressure boundary.
This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage. 3) Manual control of the block valve to unblock an isolated PORV to allow it to be used for manual control of RCS pressure and isolate a PORV with excessive leakage.
This is a function that reduces challenges to the code safety valves for overpressurization events.
- 5) Manual control of a block valve to isolate a stuck-open PORV.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The B&W pro-cess (or methad) equivalent to the inspection method described in Topical Report BAW-2045(P)-A will be used. Inservice inspection of steam generator sleeves is also required to ensure RCS integrity.
Because the sleeves intro-duce changes in the wall thickness and diameter, they reduce the sersitivity of eddy current testing, therefore, special inspection methods must be used.
A method is described in Topical Report BAW-2045(P)-A with supporting validation data that demonstrates the inspectability of the sleeve and underlying tube.
As required by NRC for licensees authorized to use this repair process, McGuire commits to validate the adequacy of any system that is used for periodic inser-vice inspections of the sleeves, and will evaluate and, as deemed appropriate by Duke Power Company, implement testing methods as better methods are developed and validated for commercial use.
McGUIRE - UNITS 1 AND 2 B 3/4 4-3 Amendment No.150 (Unit 1)
Amendment No.132 (Unit 2)
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) l The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negiigible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage - 500 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plantshutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Repair will be required for all tubes with imperfections exceeding the repair limit of 40% of the tube nominal wall thickness.
Installed sleeves with imper-fections exceeding 40% of the sleeve nominal wall thickness will be plugged.
Defective steam generator tubes can be repaired by the installation of sleeves which span the area of degradation, and serve as a replacement pressure boundary for the degraded portion of the tube, allowing the tube to remain in service. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20% of the original tube hall thickness.
For tubes with degradation below the 1
F* distance, and not degraded within the F* distance, repair is not required.
If a tube is sleeved due to degradation in the F* distance, then any defects in the tube below the sleeve will remain in service without repair.
i l
l Amendment No.150 (Unit 1)
McGUIRE - UNITS 1 and 2 B 3/4 4-3a Amendment No.132 (Unit 2)
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