ML17331A091

Proposed Tech Specs Supporting Request to Allow Interim Plugging Criteria of 1.0 Volt for Cycle 14

ML17331A091
Person / Time
Site:	Cook
Issue date:	03/10/1993
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17331A090	List: ML17331A089 ML17331A091
References
NUDOCS 9303160110
Download: ML17331A091 (12)

Text

ATTACHMENT 3 to AEP:NRC:1166G PROPOSED REVISED TECHNICAL SPECIFICATIONS PAGES PDR P, 'DR 9303i601ip 9303i0 ADQCK'50003i 5

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued Tubes in those areas where experience has indicated 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.

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:

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.

d. 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:

Cater Ins ection Results C-l 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 i 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 4-8 AMENDMENT NO. %3, ~~

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

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.

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. Seal line resistance greater than or equal to 2.27E-l ft/gpm~ and,

f. 1 GPM leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.

APPLICABILITY: MODES 1, 2, 3 and 4.~

ACTION'

~ With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours 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 />.

b. 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 />.

c ~ 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.

Amendment No. +62-466-COOK NUCLEAR PLANT - UNIT 1 3/4 4-16 Order dated April 20, 1981

\ ~

C ~

EACTOR COOLANT SYSTEM

~ASES 3 4 4 5 STEAM GENERATORS TUBE INTEGR The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integzity 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 suzveillance 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.

Insezvtce 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. Xf the secondary coolant chemistry is not maint:ained 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-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 t:he 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.

develop in service, it However, even if a defect: of similar type should will be found during scheduled inservice st:earn 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 demonst:rated the capability to reliably detect degradation that has penetrated 20X of the original tube wall thickness.

Tubes expeziencing 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.10, COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-2a AMENDMENT NO. 403, ~

EACTOR COOLAN SYST 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 uncertainti.es, 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 within 10 percent of the 10 CFR 100 guidelines. If the projected end of cycle distribution of crak indications results in primary-to-secondazy leakage greater than 120 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 120 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 LFAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operati.on 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 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 operati,on 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 permi,tting continued operation within the restrictions of the Transi.ent 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. ~~

ATTACHMENT 4 to AEP:NRC:1166G CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES

CTO COOLAN S EH SUR ILLANCE IREME S Continued

2. Tubes in those areas where experience has indicated 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 ad5acent tube shall be selected and sub]ected to a tube inspection.

c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each insezvice inspection may be sub]ected to a partial tube inspection provided:

1. 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.

H"'.

Implementation of the steam generator tube/tube supp plate interim plugging criteria for one fuel cycle (Cycle 'equires a 100't 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:

CategoO Ins ection Results C-l Less than 5t 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 1% of the total tubes inspected are defective, or between 5% and 10%

of the total tubes inspected are degraded tubes.

C-3 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. O8, 454 166

0 G C 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. I GPE mIDEETZEZED IEliEAGE, C ~ 600 gallons pez day total pz ry-to-secondazy leakage through all steam generators and 150 ons per day through any one steam generator for Fuel Cycle P

d. 10 GPE IDENTIFIED LEDGE from the Reactor Coolant System,

e. Seal line resistance greater than oz equal to 2.27E-l ft/gpm~ and,

f. 1 GPN leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.

AOTZOE'ith any PRESSURE BOUNDARY LEAKAGE; be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 houzs.

b. With any Reactor Coolant System leakage greater than any one of the abo~e limits, excluding PRESSURE BOUNDARY IKKAGE, 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 /> oz 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 />.

C ~ With any reactor coolant system pressure isolation valve(s) leakage greater than the above limit, except when:

l. 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 ~iRA 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 pzessure-within 20 .

psi of the nominal full pressure value.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-16 AMENDMENT NO. 444 166 Order dated April 20, 1981

4.4 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 tha the second-ary coolant will be maintained within those chemistry limits ound to result in negligible corrosion of the steam generator tubes. If e secondary coolant chemistry is not maintained within these paramete limits, localized corrosion may likely result in stress corrosion cracking The extent of cracking during plant operation would be limited by the imitation of steam generator tube leakage between the primary coolant syst and the secondary coolant system. The allowable primary-to-secondary le rare is 150 gallons per day per steam generator for one fuel cycle (Cycle . Axial or circumferentially oriented czacks 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 contzol and steam control.

Wastage-type defects aze unlikely with the all volatQ.e treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled insezvice steam generator tube examinations. Plugging or sleevtng 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 planes 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. ~~ 166

s'.

0 CO CO XhZZ (600 gpd Maintaining an operating total) for Fuel Cycle 1

r ge limit of will minimize 150 gpd per steam generator the potential for a large leakage event during steam line reek 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 120 gpm in the faulted loop and 150 gpd per steam generator in the intact loops, which will within 10 percent of the 10 CFR 100 guidelines.

l~t If offsite doses to the projected end of cycle distribution of crack indications results in primary-to-secondary leakage greater than 120 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 120 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.

II The Surveillance Requirements foz 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 KiEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimi ed and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Haintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the stzuctural 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 hK?HDHENT NO M 6

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