ML20116A922
| ML20116A922 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/01/1992 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20116A926 | List: |
| References | |
| NUDOCS 9210300212 | |
| Download: ML20116A922 (6) | |
Text
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'o UNITED STATES E'
.e t[,t NUCLEAR REGULATORY COMMISSION f
WASHINGTON. D. C. 20555
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SOUTHERN NVCLEAR OPERATING COMPANY. INC.
DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. NPF-2 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated October 29, 1991, as supplemented July 1, 1992, 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 license 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 amended by 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-2 is hereby amended to read as follows:
9210300212 921001 PDR ADOCK 05000348 p
PDR (2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 94
, are hereby-incorporated into the license.
Southern Nuclear 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 Elinor G. Adensam, Dire: tor Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 1, 1992 l
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- i ATTACHMENT T0-LICENSE AMENDMENT N0. 94-FACILITY OPERATING LICENSE NO. NPF-2.
DOCKET NO. 50-348 Re' place the-following.pages of the Appendix A Technical Specifications with the enclosed pages as indicated. -The revised areas are indicated by marginal-lines.
l Remove Paaes insert Paoes 3/4 4-17 3/4'4-17
- B 3/4 4-3 8 3/4 4-3 8 3/4 4-4 8 3/4.4-4 e
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'0PERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, 420 gallons per day total primary-to-secondary \\eakage through c.
all P.eam generators and 140 gallons per day through any one steam generator, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coelant System, and 4
e.
31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig.
f.
The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve shall be as specified in Table 3.4-1 at a prese we of 2235 20 psig.
6EPLICABILITY: MODES 1, 2, 3 and 4 ACTION:
a.
With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit specified in Table 3.4-1, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY withi' the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the follok.ng 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
a.
Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
Monitoring the containment air cooler condensate level system or containment atmosphere gaseous radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
FARLEY - UNIT 1 3/4 4-17 AMEDMENT NO. W, 94
s REACTOR C00LAN( SYSTEM l
_ BASES
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3/4.4.6 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 maintained. The pr: gram for inservi's inspection of steam generator tubes is based on a modification of Regelc.ory Guide 1.83, Revision 1.
Inservice inspection of steam generater tubing is essential in order to maintain surveillence 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 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 negligible 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 primary coolant system and the secondary coolant system (primary-to-secondary leakage - 140 gallons per day per j
Cracks having a primary-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. Operational leakage of this magnitude can be readily detected by existing Farley Unit 1 radiation monitors.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired.
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.
Plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness.
If a sleeved tube is found to have through wall penetration of greater than or equal to 31% fer the mechanical sleeve and 37% for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R.G.1.121 calculations with 20% added for conservatism.
The portion of the tube and the sleeve for which indications of wall degradaticn must be evaluated can be summarized as follows:
a.
Mechanical 1.
Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.
2.
Indication of tube degradation of any type including a complete guillotine breaA in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the tube be removed from service.
FARLEY-UNIT 1 B 3/4 4-3 AMENDMENT NO. 57, 72 85, 94
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REACTOR COOLANT SYSTEMI
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BASES-3/4.4?7 REACTOR COOLANT SYSTEM LEAKAGE i
- 3/4.4.7.LLEAKAGE DETECTION SYSTEMS
'The RCS_ leakage-' detection systems' required by.this specification are:
provided to monitor _and detect leakage from the Reactor Ceolant: Pressure Boundary.. These detection systems are consistent'with the recommendations of Regulatory Guide-l.45, " Reactor Cooknt Pressure. Boundary Leakage Detection Systems," May 1973.
3/4.4.7.2 OPERATIONAL LEAKAGE-
-Industry experience has shown that while a limited amount-of leakage is expected from the RCS, the unidentified portion _of this leakage can be reduced to a threshold value of less than 1 GPM..This threshold value is sufficiently low to ensure early detection of additional-leakage.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance foria limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied-to the reactor coolant pump seals exceeds-31 GPM with the -
modulating-valve in the supply line fully open at a nominal RCS pressure of-2235 psig.
This -limitation ensures that in the event of a LOCA, the safety-injection flow will not be less than assumed in the-accident analyses.
The surveillance requirements for RCS Pressure Isolation Valves. provide added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from.the RCS Pressure Isolation valves-is -IDENTIFIED LEAKAGE and will be considered a portion of the allowed limit.
The total steam generator tube leakage limit of 420 gallons.per day l
for all steam generators ensures that the dosage contribution from'the tube leakage will be: limited'to a small fraction of Part 100 limits in-the event of either a steam generator tube rupture ar steam line break. A1 GPM limit is consistent with the assumptions u.ad in-the analysis _ of these.
accidents.
The 140 gallons per day leakage limit per steam generator l
ensures that steam generator tube integrity is maintained in the event of a main steam _line rupture or under LOCA conditions.
PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable =since it may be indicative of an impending gross failure of the pressure boundary.-
Therefore, the presence of. any PRESSURE BOUNDARY LEAKAGE requires the unit -
to be promptly placed in COLD SHUTDOWN.
FARLEY-UNIT 1 8 3/4 4-4 AMENDMENT-NO. 50, 94
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