ML17331B313
| ML17331B313 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/14/1994 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17331B314 | List: |
| References | |
| NUDOCS 9403180255 | |
| Download: ML17331B313 (7) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 2OSSS-OOOI INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C.
COOK NUCLEAR PLANT UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 177 License No.
DPR-58 l.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company (the licensee) dated February 15,
- 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 set forth in 10 CFR-Chapter I;
B.
The faci]ity 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 conrpliance with the Commission's regulations; 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.
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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.
DPR-58 is hereby amended to read as follows:
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No. f77, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 14, 1994 Ledyard B. Marsh, Director Project Directorate III-I Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT NO. 177 TO FACILITY OPERATING LICENSE NO.
DPR-58 DOCKET NO. 50-315 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
REMOVE 3/4 4-8 3/4 4-10 3/4 4-11 3/4 4-12 INSERT 3/4 4-8 3/4 4-10 3/4 4-11 3/4 4-12
REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued 2 ~
3 ~
Tubes in those areas where experience has indi.cated potential problems.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
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d.
In addition to the sample required in 4.4.5.2.b.l through 3, all tubes which have had the F* criteria applied will be inspected in the roll expanded region.
The roll expanded region of these tubes may be excluded from the requirements of 4.4.5.2.b.l.
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:
The tubes selected for the samples include the tubes from those areas of the tube sheet array where tubes wi.th imperfections were previously found.
2 ~
The inspections include those portions of the tubes where imperfections were previously found.
e.
Implementation of the steam generator tube/tube support plate interim plugging criteria for one fuel cycle (Cycle 13) requires a
100%
bobbin coil inspection for hot leg tube support plate intersecti.ons and cold leg intersections down to the lowest cold leg tube support plate with known outer diameter stress corrosion cracking (ODSCC) indications.
The results of each sample inspection shall be classified into one of the following three categories:
Cate caCor r C-1 Ins ection Results Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 C-3 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
COOK NUCLEAR PLANT UNIT 1 3/4 4-8 AMENDMENT NO. 98/ +5+/ +66 $
177
ACTOR COOLANT SYS EM SURVEILLANCE RE UIREMENTS Continued 4.4'.4 cce tance Criter a a.
As used in this Specificationc contour of a tube or sleeve from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
2 ~
general corrosion occurring on either inside or outside of a tube or sleeve.
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4.
5.
6.
7 e De raded Tube or sleeve means an imperfection greater than or equal to 20%
of the nominal wall thickness caused by degradation.
Percent De radatio means the amount of the tube wall thickness affected or removed by degradation.
defect means an imperfection of such sevexity that it exceeds the repair limit.
Re air Plu in Limit means the imperfection depth at or beyond which the tube or sleeved tube shall be repaired or removed from service.
Any tube which, upon inspection, exhibits tube wall degradation of 40 percent or more of the nominal tube wall thickness shall be plugged or repaired prior to returning the steam generator to service.
This definition does not apply to the portion of the tube in the tubesheet below the F+ distance for F* tubes.
Any sleeve which, upon inspection, exhibits wall degradation of 29 percent or more of the nominal wall thickness shall be plugged prior to returning the steam generator to service.
In addition, any sleeve exhibiting any measurable wall loss in sleeve expansion transition or weld zones shall be plugged.
This definition does not apply for tubes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates.
See 4.4.5.4.a.l0 for the plugging limit for use within the thickness of the tube support plate.
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 4.4.5.3.c, above.
8.
or sleeve from the point of entry (hot leg side) completely COOK NUCLEAR PLANT - UNIT 1 3/4 4-10 AMENDMENT NO. 98, 45+, 466, 177
REACTOR COOLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued around the U-bend to the top support of the cold leg.
For a
tube in which the tube support plate elevation interim plugging limit has been applied, the inspection will include all the hot leg intersections and all cold leg intersections down to, at
- least, the level of the last crack indication.
9.
~sleevin a tube is permitted only in areas where the sleeve spans the tubesheet area and whose lower joint is at the primary fluid tubesheet face.
10 ~ The Tube Su ort Plate I terim Plu in Criteria is used for ciisposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates.
For application of the tube support plate interim plugging limit, the tube '
disposition for continued service will be based upon standard bobbin probe signal amplitude.
The plant-specific guidelines used for a3.1 inspections shall be amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals arith respect to the above voltage/depth parameters.
Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in the Donald C.
Cook Nuclear Plant Unit 1 steam generator inspections for consistent voltage normalization.
A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.0 volt, regardless of the depth of tube wall penetration, if, as a result, the projected end-of-cycle distribution of crack indications is verified to result in primary-to>>secondary leakage less than 1
gpm in the faulted loop during a postulated steam line break event.
The methodology for calculating expected leak rates from the projected crack distribution must be consistent with WCAP-13187, Rev.
0.
A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 1.0 volt except as noted in 4.4.5.4.a.10.3 belo~.
A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than or equal to 4.0 volts if a rotating pancake probe inspection does not detect degradation.
Indications of degradation with a bobbin coil signal amplitude greater than 4.0 volts will be plugged or repaired.
- 11. F* Distance is the distance from the bottom of the hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.11 inches (not including eddy current uncertainty).
12 F> Tube is a tube with degradation, below the F* distance, equal to or greater than 40%,
and not degraded (i.e.,
no indications of cracking) within the F* distance.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-11 AMENDMENT NO. 9By kS+/ 466 p
177
REACTOR COOLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued The steam generator shall be determined OPERABLE after completing the corresponding actions (plugging or sleeving all tubes exceeding the repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.
Ce Steam generator tube repairs may be made in accordance with the methods described in either WCAP-12623 or.CEN-313-P.
4.4 5
5 R~eerte a ~
b.
Following each inservice inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.
The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed.
This report shall include:
1.
Number and extent of tubes inspected.
2 ~
Location and percent of wall-thickness penetration for each indication of an imperfection.
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3.
Identification of tubes plugged or sleeved.
Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation.
The written followup of this report shall provide a
description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
d.
The results of inspections performed under 4.4.5.2 for all tubes in which the tube support plate interim plugging criteria has been applied or that have defects below the F+ distance and were not plugged shall be reported to the Commission within 15 days following the inspection.
The report shall include:
1.
Listing of applicable tubes.
2.
Location (applicable intersections per tube) and extent of degradation (voltage).
The results of steam line break leakage analysis performed under T/S 4.4.5.4.a.10 will be reported to the Commission prior to restart for Cycle 13.
COOK NUCLEAR PLANT UNIT 1 3/4 4-12 AMENDMENT NO. +&&, +66, 177