ML20113G680
| ML20113G680 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/01/1992 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20113G681 | List: |
| References | |
| NUDOCS 9204160088 | |
| Download: ML20113G680 (10) | |
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NUCLEAR REGULATORY CGMMISSION
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s, SOUTHERN NUCLEAR OPERATING COMPW 1 J L DOCKET N0. 50-351 JDSEPH M. FARLEY NUCLEAR PLANT. UNIT 2 AMENDMENT 10 FACILITY OPERATING LICENS1 Amendment No. 87 License No. NPF-8 1.
The Nuclear Regulatory Commission (th ommission) has found that:
A.
The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated February 20, 1992, as supplemented on March 27, 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..he health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicabla requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical l
Specifications, as indicated in the attachment to this license amendment: and paragraph 2.C.(2) of Facility Operating License No.
NPF-8 is hereby amended to read as follows:
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9204160088 920401 PDR ADDCK 05000364 P
PDR i
. (2) Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 87, are hereby incorporated in 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 iW Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of 'ssuance: April 1, 1992 l
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ATTACPMENT TO LICENSE AMENDMENT NO. 87 ffl]LITY _0PERATING LICENSE N.Q, NPF-8
@CKET NO. 50-364 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Remove Paaes Insert Paaes 3/4 4-12 3/4 4-12 3/4 4-12a 3/4 4-17 3/4 4-17 3/4 4-17a B3/4 4-3 83/4 4-3 83/4 4-3a 83/4 4-3a 83/4 4-3b l
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RE1.CTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.4 Accectance Criteria a.
As used in this Specification:
1.
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication orawings or specifications.
Eddy-current testing indications below 20% of the cominal wall thickness, if detectable, may be considered as imperfections.
2.
Dearadation means a service-induced cracking,
-wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.
3.
Dearaded Tube means a tube,
including the sleeve if the tube has been repaired, that contains-imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
4.
W Dearadation means the percentage of the tube or sleeve wall-thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the plugging or repair limit.
A tube or sleeve containing a defect is defective.
6.
Plulaina or Repair limii means the imperfection depth at or beyono whia the tube shall be repaired (i.e., sleeved) or removed from service by plugging and is greater than' or equal to 40% of the nominal tube wall thickness.
This definition de:
St apply to the area of the tubesheet region below the F* d. n :e in the F*
tubes.
For a tube that has been sleeved with a anical joint sleeve, through wall penetration of greater than or equal to 31%
of sleeve nominal wall thickness in the sleeve requires the tube to be removed from' service by plugging.
For a tube that has been-sleeved with a welded joint sleeve. -through wall penetration greater than or equal to 37X of sleeve nominal wall thickness in the sleeve between the weld joints requires the tube to be removed from service by plugging. At tube supm rt' plate intersections,
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the repair limit for the Ninth Operatie3 Cycle is based on maintaining steam generator tube serviceability as described below:
a.
An eddy current examination using a bobbin probe of 100% of the hot and cold leg steam generator tube support. plate-intersections will be-performed-for tubes in service..
b.
Degradation within the bounds of the tube support plate with bobbin voltage less than or equal-to 1.0 volt will be allowed to remain in service.
c.
Degradation within the bounds of the tube support plate with-a bobbin voltage greater-than 1.0 volts will be' repaired or.
plugged except as noted in 4.4.6.4.a.6.d below.
FARLEY-UNIT 2 3/4 4-12 AMENDMENT rio. 63, H 78 87'
' REACTOR COOLANT 5(STEM SVRVEILLANCE REQUIREMENTS (Continued)
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d.
Indications of potential degradation within the Dounds of the tube support plate with a bobbin voltage greater than 1.0 volt but less than or equal to 3.6 volts may remain in service if a rotating pancake coil probe (RPC) inspection does not detect degradation.
Indications of degradation with a bobbin voltage greater than 3.6 volts will be plugged or repaired.
7.
Unserviceable describes the condition of a tube or sleeve if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4,4.6.3.c, above.
8.
Tulie Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the V-bend to the top support of the cold leg.
For a tube that has been repaired by sleeving, the tube inspection should include the sleeved portion of the tube.
9.
Tube Reoair refers to mechanical sleeving, as described by Westinghouse report WCAP-ll178, Rev. 1, or laser welded sleeving, as described by Westinghouse report WCAP-12672, which is used to maintain a tube in service or return a tube to service.
This includes the removal of plugs that were installed as a corrective or preventive measure.
FARLEY-UNIT 2 3/4 4-12a AMENDMENT N0. 87 l
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, c.
For the Ninth Operatino Cycle only. orimary-to-secondary leakaae-throuah all steam cenerators shall te limited to 450 callons Der day and 150 callons Der day throuah any one steam gsnerator.
For subsequent cycles, 1 GPM total primary-to-secondary l
leakage through all steam generators and 500 gallons per day -
through any one stea, generator, d.
10 GPM UNIDENTIFIED LEAKAGE from the Reactor Coolant System, and 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 pressure of 2235 1 20 psig.
APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:
a.
With any PRESSURE BOUNDARY 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 ^1LD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, i
b.
With any Reactor Coolant System leakage greater than any one-of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, l-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 l
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 l
SHUTDOWN within the following 30'huurs, 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. hours by use.of at least-two closed manual or deactivated automatic valves, or be in at least HOT-STANDBY-within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s-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 />.
FARLEY-UNIT 2-3/4 4-17 AMENDMENT NO. 47, 87 l
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'0VEILLANCE REQUIREMENTS j
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 />, t
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 />.
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FARLEY-UNIT 2 3/4 4-17a AMENDMENT NO.
J
- - - - -. - - = -.-
REAC:0R COOLANT SYSTEM BASES 3!4.4.6 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this pnrtion of the RCS will be maintained.
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 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 seconoary coolant system (primary-to-secondary leakage. 500 gallons per day per steam geo rator).
Cracks having a primary-to-secondary leakage less than this lirrit 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 primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation moniturs of steam generator blowdown.
Leakage in excess of this limit wil' requirn plant shutdown and an unscheduled inspection, during which tb leaking tubes will be located and plugged or repaired.
For the Ninth Operating Cycle only, the repair limit for tubes with flaw indications contained within the bounds of a tuae support plate has been provided to the NRC in Southern Nuclear Operating Company letter dated February 20, 1992.
The repair limit is based on the analysis contained in WCAP-12871, Revision 2. "J. M. Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates." The application of this criteria is based on limiting primary-to-secondary leakage during a steam line break to less thaa 1 gallon per minute.
Primary-to-secondary leakage during this cycle only is limited to 150 gallons per day per steam generator during normal operation.
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% for the mechanical sleeve and 37% for the laser welded sleeve cf 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 degradation must be evaluated can be summarized as follows:
FARLEY-UNIT 2 B3/4 4-3 AMENDMENT N0.
6, H, 78 87
REACTOR COOLANT 5151FJ BASES 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 break in the tube between the bottom of the upper joint and the to) of the lower roll expansion does not require that the tube se removed from service.
3.
The tube plugging limit continues to apply to the portion of the tube in the entire upper joint region and in the lower roll expansion. As noted above the sleeve plugging limit applies to these areas also.
4.
The tube plugging limit continues to apply to that portion of the tube above the top of the upper joint, b.
Laser Welded 1.
Indications of degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.
2.
Indication of tube degradation of any type including a complete break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed from service.
3.
At the weld joint, degradation must be evaluated in both the sleeve and tube.
4.
In a joint with more thar one weld, the weld closest to the end of the sleeve represents the joint to be inspected and the limit of the sleeve inspection.
5.
The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint.
F* tubes do not have to be plugged or repaired provided the remainder of the tube within the tubesheet that is above the F* distance is not degraded.
The F* distance is equal to 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transitio.., whichever is lower in elevation.
included in this distance is an allowance of 0.25 inch for eddy current elevation measurement uncertainty.
Steam generator tube inspections cf operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20%
of the original tube wall thickness.
FARLEY-UNIT 2 B3/4 4-3a AMENDMENT NO. 63, 64.
78, 87
-. = - _._
REACTOR COOLANT SYSTEM BASES Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to 10 CFR 50.73 prior to resumption of plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision to the Technical Specifications, if necessary.
(
FARLEY-UNIl 2 B3/4 4-3b AMENDMENT NO. 87
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