ML17331B309
| ML17331B309 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/15/1994 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17331B310 | List: |
| References | |
| NUDOCS 9403180233 | |
| Download: ML17331B309 (13) | |
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e UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C.
COOK NUCLEAR PLANT UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 178 License No. DPR-58 l.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
B.
C.
D.
E.
The application for amendment by Indiana Michigan Power Company (the licensee) dated December 15,
- 1993, as supplemented February 15,
- 1994, and February 24, 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; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; 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 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.
9403180233 940315 PDR ADOCK 05000315 PDR
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.
178
, 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:
Narch 15, 1994 Ledyard B. Harsh, Director Project Directorate III-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
TTACHMENT TO LICENSE AMENDMENT NO.
178 TO FACILITY OPERATIAG 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 4-11 3/4 4-12 3/4 4-16 B 3/4 4-2a B 3/4 4-4 INSERT 3/4 4-8 3/4 4-11 3/4 4-12 3/4 4-16 B 3/4 4-2a B 3/4 4-4
REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS 2.
Tubes in those areas where experience has indicated potential problems.
3
~
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.
In addition to the sample required in 4.4.5.2.b.l through 3, all tubes which have had the W criteria applied will be inspected in the zoll expanded region.
The roll expanded region of these tubes may be excluded from the requirements of 4.4.5.2.b.l.
d.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each insezvice 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 with 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 14) requires a
100X bobbin coil inspection for hot leg tube support plate intersections 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 or Ins ection Results C-1 Less than 5X of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1X of the total tubes inspected are defective, or between 5X and 10X of the total tubes inspected are degraded tubes.
C-3 More than lOX of the total tubes inspected are degraded tubes or more than 1X of the inspected tubes are defective.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-8 AMENDMENT NO. ~, ~, ~,
178
REACTOR COOLANT SYSTEM 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 Interim Plu in Criteria is used for disposition 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's disposition for continued service will be based upon standard bobbin probe signal amplitude.
The plant-specific guidelines used for all inspections shall be amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the above voltage/depth parameters.
Pending incorporation of the voltage verification requirements 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 2.0 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 12.6 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, and as prescribed in draft NUREG-1477.
2.
A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 2.0 volt except as noted in 4.4.5.4.a.10.3 below.
A tube can remain in service with a bobbin coil signal amplitude greater than 2.0 volt but less than or equal to 3.6 volts if a rotating pancake probe inspection does not detect degradation.
Indications of degradation with a bobbin coil signal amplitude greater than 3.6 volts will be plugged or repaired.
I 11 '
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, pe Tube is a tube with degradation, below the pe distance, equal to or greater than 40',
and not degraded (i.e.,
no indications of cracking) within the W distance.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-11 AMENDMENT NO. ~, ~, +7+,
178
ACTOR COO SYS EMS SURVEILLANCE RE UIREMENTS Continued b.
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.
C ~
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 e orts 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.
C ~
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 14.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-12 AMENDMENT NO. 4&3 g hap 178
ACTOR COOLANT SYSTEM 0 ERATIONAL LEAKAGE IMITING CONDITION FOR OPERATION
- 3. 4. 6. 2 Reactor Coolant System leakage shall be limited to c a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, C ~
600 gallons per day total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator for Fuel Cycle 14, I
d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, e.
Seal line resistance greater than or equal to 2.27E-1 ft/gpm~ and, 1
GPM leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.
APPLICABILITY MODES 1 ~
2 g 3 and 4 ~ **
~CT TON:
a.
b.
CN 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 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 />.
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 />.
With any reactor coolant system pressure isolation valve(s) leakage greater than the above limit, except when:
The leakage is less than or equal to 5.0 gpm, and 2 ~
The most recent measured leakage does not exceed the previous measured leakage>>
by an amount that reduces the To satisfy ALARA requirements, measured leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
specification 3.4.6.2.e is applicable with average pressure within 20 psi of the nominal full pressure value.
COOK NUCLEAR PLANT UNIT 1 3/4 4-16 AMENDMENT NO. 46&, 466,
] 78 Order dated April 20, 1981
~SES 3 4.4.5 STEAM GENERATORS TUBE I TEG ITY 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 program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision l.
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 second-ary 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 parameter 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.
The al lowabe primary-to-secondary leak rate is 150 gallons per day per steam generator for one fuel cycle (Cycle 14).
Axial or circumferentially oriented 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.
Leakage in excess of this limit will require plant shutdown and an inspection, during which the leaking tubes will be located and plugged or repaired.
A steam generator while undergoing crevice flushing in Mode 4 is available for decay heat removal and is operable/operating upon reinstatement of auxiliary or main feed flow control and steam control.
Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.
- However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the repair limit which is defined in Specification
- 4. 4. 5. 4. a.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired by the criteria of 4.4.5.4.a.10.
COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-2a AMENDMENT NO. 408, AS+, 466, 1 78
REACTOR COOLANT S STEM BASES Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) for Fuel Cycle 14 will minimize the potential for a large leakage event during steam line break under LOCA conditions.
Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 12.6 gpm which willlimit the calculated offsite doses to within 10 percent of 10 CFR 100 guidelines.
Leakage in the intact loops is limited to 150 gpd. If the projected end of cycle distribution of crack indications results in primary-to-secondary leakage greater than 12.6 gpm in the faulted loop during a postulated steam line break
- event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 12.6 gpm.
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Should PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.
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 as a portion of the allowed limit.
3 4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.
The associated effects of exceeding the oxygen,
- chloride, and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-4 AMENDMENT NO. $8, 466, 178