ML20137D675
| ML20137D675 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/24/1997 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137D682 | List: |
| References | |
| NUDOCS 9703260264 | |
| Download: ML20137D675 (11) | |
Text
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UNITED STATES g
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NUCLEAR REGULATORY COMMISSION o
WASHINGTON. D.C. 30666 4 001 i
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SOUTHERN NUCLEAR OPERATING COMPANY. INC.
ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.124 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 December 26, 1996, as supplemented by letters dated February 6, March 7, and March 21, 1997, 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 Co;.1 mission's regulations; j
4 D.
The issuance of this license amendment will not be inimical to the l
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-8 is hereby amended to read as follows:
9703260264 970324 PDR ADOCK 05000348 P
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Technical Specifications l
l The Technical Specifications contained in Appendices A and B, as j
revised through Amendment No. 124, are hereby incorporated in the l
license. Southern Nuclear shall operate the facility in l
accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within'30 days of issuance.
j i
FOR THE NUCLEAR REGULATORY COP #ilSSION (shh--A).
hr Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
4 Changes to the Technical Specifications Date of Issuance:
March 24, 1997 I
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A i
ATTACHMENT TO LICENSE AMENDMENT NO.124 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 i
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 Paoes 3/4 4-10 3/4 4-10 3/4 4-12 3/4 4-12 3/4 4-12a 3/4 4-12a 3/4 4-12b 3/4 4-13 3/4 4-13 B 3/4 4-3 B 3/4 4-3 j
B 3/4 4-3a B 3/4 4-3a B 3/4 4-3b l
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l 4
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4 4
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SURVEILLANCE REQUIREMENTS (Continued) 1.
All nonplugged tubes that previously had detectable wall penetrations greater than 20%.
2.
Tubes in those areas where experience has indicated potential problems.
3.
At least 3% of the total number of sleeved tubes in all three steam generators or all of the sleeved tubes in the generator chosen for the inspection program, whichever is less. These inspections will include both the tube and.the sleeve.
l-4.
A tube inspection (pursuant to Specification 4.4.6.4.a.8) shall be performed on each selected tube.
If any sslected tube does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
l 5.
Indications left in service as a result of application of the tube suppcrt plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.
c.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections were previously found.
d.
Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.
The results of each sample inspection shall be classified into one of the following three categories:
i-l FARLEY-UNIT 1-3/4 4-10 AMENDMENT NO.
124
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e SURVEILLANCE REQUIREMENTS (Continued) l 4.4.6.4 Acceptance Cr'iteria j
l a.
As used in this Specification:
i 1.
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that l
required by fabrication drawings or specifications.
?
Eddy-current testing indications below 20% of the l
nominal wall thickness, if detectable, may be i
considered as imperfections.
a 2.
Degradation ~means a service-induced cracking, wastage, wear or general corrosion occurring on either inside
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or outside of a tube or sleeve.
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3.
Degraded Tube means a tube, including the sleeve if i
the tube has been repaired, that contains imperfections greater than or equal to 20% of the i
nominal wall thickness caused by degradation.
i 4.
% Degradation means the percentage of the tube or i
sleeve wall thickness affected or removed by j
degradation. -
I a
I 5.
Defect means an imperfection of such severity that it j
exceeds the plugging or repair limit. A tube or sleeve containing a defect is defective.
l 4
l 6.
Plugging or Repair Limit means the imperfection depth I
at or beyond which the tube shall be repaired (i.e.,
)
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sleeved) or removed from service by plugging and is
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greater than or equal to 40% of the nominal tube wall J
i thickness.
For a tube that has been sleeved with a mechanical joint sleeve, through wall penetration of
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j greater than or equal to 31% of sleeve nominal wall l
1 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 l
penetration greater than or equal to 37% of sleeve nominal well thickness in the sleeve between the weld I
joints requires the tube to be removed from service by plugging. This definition-does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.
Refer to 4.4.6.4.a.ll for the repair limit applicable to these intersections.
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.
FARLEY-UNIT 1 3/4 4-12 AMENDMENT No. 124 i
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-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 Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-11178, 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.
10.
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
11.
Tube Support Plate Repair Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:
a.
Steam generator tubes, whose degradation is i
attributed to outside diameter stress corrosion j
cracking within the bounds of the tube support plate with bobbin voltage less than or equal to the lower voltage repair limit [2.0 volts), will be allowed to remain in service.
b.
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair lindt (2.0 volts), will be repaired or plugged except as noted in 4.4.6.4.a.11.c below.
FARLEY-UNIT 1 3/4 4-12a AMENDMENT NO. 124
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Csntinusd)
Steam generator tubeIs, with indications of c.
potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit [2.0 volts), but less than or equal to the upper voltage repair limit *, may remain in service if a rotating probe inspection does not
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detect degradation.
Steam generator tubes, with indications of outside diameter stress corrosion j
cracking degradation with a bobbin voltage greater than the upper voltage repair limit *,
will be plugged or repaired.
d.
If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 4.4.6.4.a.11.a, 4.4.6.4.a.ll.b, and 4.4.6.4.a.11.c.
The mid-cycle repair limits are determined from the following equations:
Vn Vmu= 1.0 + NDE + Gr [ CL-At )
CL Vee.u=Vmg-'[Vog-Vg,g) [ CL-At }
CL where:
upper voltage repair limit Vuu
=
lower voltage repair limit V,u
=
mid-cycle upper voltage repair limit Vmn
=
based on time into cycle mid-cycle lower voltage repair limit J
Vnua
=
based on Vmu and time into cycle j
length of time since last scheduled At
=
inspection during which Vug and Vp; were implemented j
cycle length (the time between two j
CL
=
scheduled steam generator l
inspections)
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structural limit voltage Vn
=
average growth rate per cycle length Gr
=
95-percent cumulative probability NDE
=
allowance for nondestructive examination uncertainty (i.e.,
a value of 20-percent has been
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approved by NRC)
Implementation of these mid-cycle repair lindts should follow the same approach as in TS 4.4.6.4.a.11.a, 4.4.6.4.a.11.b, and 4.4.6.4.a.ll.c.
The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.
FARLEY-UNIT 1 3/4 4-12b AMENDMENT NO.124
SURVEILLANCE REQUIREMENTS (Cantinutd) b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair of all tubes exceeding the plugging or repair lindt) required by Table 4.4-2.
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4.4.6.5 Repor!
4 a.
Following each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam i
generator shall be reported to the Commission within 15 days j
of the completion of the plugging or repair effort.
J b.
The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in 4
a special Report pursuant to Specification 6.9.2 within 12 1
months following the completion of the inspection. This I
special Report shall include:
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1.
Number and extent of tubes and sleeves inspected.
l 2.
Location and percent of wall-thickness penetration for j
each indication of an imperfection.
3.
Identification of tubes plugged or repaired.
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c.
Results of steam generator tube inspections which fall into Category C-3 shall be considered a REPORTABLE EVENT and l
shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. The written report shall provide a description of investigations conducted to determine the i
j cause of the tube degradation and corrective measures taken to prevent recurrence.
l d.
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior 1
to returning the steam generators to service (Mode 4) should any of the following conditions arise:
l 1.
If estimated leakage based on the projected end-of-j cycle (or if not practical, using the actual measured i
j end-of-cycle) voltage distribution exceeds the leak l
limit (determined from the licensing basis dose calculation for the postulated main steam line break)
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for the next operating cycle.
l 2.
If circumferential crack-like indications are detected
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4 at the tube support plate intersections.
3.
If indications are identified that extend beyond the confines of the tube support plate.
4.
If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
i 5.
If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10'#. notify the NRC and provide an assessment of the safety significance of the occurrence.
FARLEY-UNIT 1 3/4 4-13 AMENDMENT NO.124 i
REACTOR COOLANT SYSTEM BASES I
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3/4.4.6 STEAM GENERATORS I
The Surveillance Requirements for inspection of the steam generator tubes i
ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is 3'
based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice j
inspection of steam generator tubing is essential in order to maintain
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surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, i
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
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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 steam generator).
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 I 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.
The voltage-based repair limits of 4.4.6 implement the guidance in GL 95-05 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube j
support plate intersections. The voltage-based repair limits are not j
applicable to other forms of SG tube degradation nor are they applicable to 1
ODSCC that occurs at other locations within the SG.
Additionally, the repair j
criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no significant cracks extending outside the thickness of the support plate.
Refer to GL 95-05 for additional description of the degradation morphology.
1 i
Implementation of 4.4.6 requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural lindt (which is then implemented by this surveillance).
FARLEY-UNIT 1 B 3/4 4-3 AMENDMENT No.124
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BASES 1
j The voltage structural limit is the voltage from the burst pressure / bobbin i
voltage correlation at the 95-percent prediction interval curve reduced to i
account for the lower 95/95-percent tolerance for tubing material properties l
at 650 *F (i.e.,' the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty. The upper voltage repair limit, j
Vom, is detennined from the structural voltage limit by applying the following equation:
Vuu = Vet - Vr., - Van i
where Va, represents the allowance for flaw growth between inspections and Van l
- represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to j
determine the voltage repair limit is contained in GL 95-05.
The mid-cycle equation in 4.4.6.4.a.11.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube j
support plates.
4.4.6.5 implements several reporting requirements recommended.by GL 95-05 for situations in which the NRC wants to be notified prior to returning the SGs to l
service.
For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage =
i distribution rather than the projected end-of-cycle voltage distribution l
(refer to GL 95-05 for more information) when it is not practical to complete l-these calculations using the projected EOC voltage distributions prior to i
returning the SGs to service. Note that if leakage and conditional burst l
probability were calculated using the measured EOC voltage distribution for j
the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting criteria, j
then the results of the projected EOC voltage distribution should be provided I
per the GL section 6.b(c) criteria.
l 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.
4 Plugging or repair will be required for all tubes with imperfections exceeding
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40% of the tube nominal wall thickness. If a sleeved tube is found to have i
through wall penetration of greater than or equal to 314 for the mechanical sleeve and 374 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.
i' l.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:-
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i FARLEY-UNIT 1 B 3/4 4-3a AMENDMENT NO.124 1
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BASES i
- a. Mechanical 1.
Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.
l 2.
Indication of tube degradation of any type including a complete l
gui13stine break in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the a
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tube be removed from service.
2 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 j
expansion. As noted above, the sleeve plugging lLnit 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 i
4 1.
Indications of degradation in the length of the sleeve between the
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weld joints must be evaluated against the sleeve plugging lindt.
t i
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 than one weld, the weld closest to the end of the sleeve represents the joint to be inspected and the lindt 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.
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 wall thickness.
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 CrR 50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and nay result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision to the Technical Specifications, if necessary.
FARLEY-UNIT 1 B 3/4 4-3b AMENDMENT NO.124 l
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