ML17331B141
| ML17331B141 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/09/1993 |
| From: | Pitterle T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17331B140 | List: |
| References | |
| NSD-TAP-3052, NUDOCS 9312270245 | |
| Download: ML17331B141 (4) | |
Text
REACTOR COOLANT SYSTEM SURVEILLANCE RE UIRENENTS Cont:inued 2., Tubes in those areas where experience has indicated potent:ial problems.
3.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each select:ed 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 t:o a tube inspection.
c.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspecti.on may be subjected to a partial tube inspection provided:
1.
The tubes selected f'r the samples include the tubes from
'hose areas of the tube, sheet arxay where tubes with
'mperfect:ions were previously found.
d.
2.
The inspections include those portions of the tubes where imperfections were previously found.
Implementation of the steam generator tube/tube support plate interim plugging criteria for one fuel cycle (Cycle 14) requires a
10OX 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 cat:egories:
~Cate ~or Ins ection Results C-1 Less than 5X of the tot:al 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 10X 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'-8 AMENDMENT NO. 8S,~~
9312270245 931215 PDR ADOCK 05000315 P
REACTOR COOLANT SYSTEMS SURVEILLANCE RE UIREMENTS Cont:inued 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 int:ersections and all cold leg intersections down to, at least, the level of the last crack indicat:ion.
9.
~aleevin 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 cont:inued 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 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.
1.
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 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 1.0 volt except as noted in 4.4.5.4.a.10.3 below.
3.
A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than or equal to 3.1 volts if a'otating pancake'robe inspect:ion does not detect degradation.
Indications of degradation with a bobbin coil signal amplitude greater than 3.1 volts will be plugged or repaired.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-11 AMENDMENT NO. 08,. ~,
EACTOR COOLANT SYSTEM ASES 3 4 4 5 STEAM GENERATORS TUBE INTEGRI The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS willbe 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 pzogressive degradation due to design, manufacturing errors, or insezvice conditions that lead to corrosion.
Inservice inspection of steam genezator 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 allowable primary-to-secondazy 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 limitwill 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 willbe 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 20X'of the original tube wall thickness.
Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support are plugged or repaired by the criteria of 4.4.5.4.a.lO.
COOK NUCLEAR P~T - UNIT 1 B 3/4 4-2a AMENDMENT NO.
REACTOR COOLANT SYSTEM BASES Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) for Fuel Cycle 14 villminimize the potential for a large leakage event during steam line break under LOCA conditions.
Based on the HDE 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 120 gpm in the faulted loop and 150 gpd per steam generator in the intact loops, which will limit offsite doses to vithin 10 percent of the 10 CFR 100 guidelines.
If the projected end of cycle distribution of crak indications z'esults in primary-to-secondary leakage greater Chan 120 gpm in the faulted loop during a postulated sCeam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to belov 120 gpm.
PRESSURE.BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure o'K 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 L1BKAGE and will be considered as a portion of the allowed limit.
3 4 4 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 LimiCs 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'ignificant 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. ~+6&