ML17332A556

Proposed Tech Specs to Incorporate 2.0 Volt Interim Steam Generator Tube Support Plate Plugging Criterion for Fuel Cycle 15

ML17332A556
Person / Time
Site:	Cook
Issue date:	02/03/1995
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17332A555	List: ML17332A553 ML17332A556
References
NUDOCS 9502130202
Download: ML17332A556 (19)

Text

ATTAQiMENT 2 TO AEP'NRC:1166/

EXISTING TECHNICAL SPECIFICATION PAGES MARKED TO REFLECT PROPOSED CHANGES 9502130202 950203 PDR ADQCK 050003i5 P PDR

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~ i EACTOR COOLANT SYSTEM SURVEILLANCE REQUIRE."KNTS

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 tube does not permit the passage of the eddy current probe for any selected a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

c ~ 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 roll 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 inse'rvice 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.

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e. Implementation of the steam generator tube/tube s pport plate interim plugging criteria for one fuel cycle (Cycle 4 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:

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 lOX 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 4-8 AMENDMENT NO. ~, ~, ~, 178

REACTOR COOPT SYS E'HS SURVE'ZLLANC RE UTR:-.M:-NTS Cont'ued 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.

Ce Steam gene ator 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 ac Following each inservice inspection of steam generator tubes, are any tubes requiring plugging or sleeving, the number of tubes if there plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

b. The complete results of the steam gene ator 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.

3. Identification of tubes plugged or sleeved.

C ~ Results of steam generator tube inspections which fall into Category C-3 and equire prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant ope ation. The written followup of this report shall provide a descript'on of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

de 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:

Listing of applicable tubes.

2~ Location (applicable intersections per tube) and extent of degradation (voltage).

e. The results of steam line break leakage analysis performed under 4.4.5.4.a.lO will be reported to the Commission prior to restart T/S for Cycle~.

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COOK NUCLEAR PLANT - UNIT 1 3/4 4-12 KKNDHENT NO. +S&g 466p +ii p 178

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REACTOR COOLANT SYSTEM O>ERATIONAL LEAKAGE IMITING CONDITION FOR OPERATION Reactor Coolant System leakage shall be limited to:

3.4 6.2

a. 'o PRESSURE BOUNDARY LEAKAGEt

b. 1 GPM UNZDENTZFZED LEAKAGEt Co 600 gallons per day total primary-to-secondary..leakage,.through all steam generators and 150 allons per day through any one steam generator for Fuel Cycle 4 I 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 lg 2, 3 and 4. <<<<

~CTION:

a. 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 />.

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 />.

Co 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 o f 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.

AMENDYZNT NO. 46&, +66, ].78 COOK NUCLEAR PLANT UNIT 1 3/4 4-16 Order dated April 20, 1981

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REACTOR COO~T SYS BASES 3 4.4.5 S EAM GENERATORS TUBE INTEGRITY 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 af steam generator tubes i.s based on a modification of Regulatory Guide 1-83, Revision 1. Znservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions cif 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. Znservice 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 i.n a manner such that the second-ary coolant will be maintained within those chemistry limits found to r negligible corrosion of the steam genezator tubes. Zf the s ry coolant chemistry i.s not maintained within these parameter limit calized corrosion may likely result in stress corrosion cracking. The ent of cracking duri.ng plant operation would be limited by the limi. on of steam generator tube leakage between the primary coolant system e secondary coolant system. The allowabe primary-to-secondary leak generator for one fuel cycle (Cycle t s 150 gallons per day per steam

. Axial or circumferent'ally oriented c acks 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 gene ator while undergoing crevice flushing in Made 4 is available for decay heat removal and is operable/operating upon reinstatement of auxiliary or main feed flow control and s earn control.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of seconda y service, coolant.

it However, even will be found if a defect during scheduled of similar type should develop in inservice steam generator tube examinations. Plugging or sleeving will be required for all tubes with imperfec ions 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 thi.ckness.

Tubes experiencing outer diameter stress corrosion cracking withi.n 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 AYENDMENT NO. 403, @Sees k66 s 178

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REACTOR COOLANT SYSTEM BASES Maintaining an operati leakage limit, of 150 gpd per steam generator (600 gpd total) for Fuel cycle 4 will minimize the potential for a large leakage event, during steam line zeak under LocA conditions. Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous i.nspection, the expected leak rate following a steam line rupture is limited to below 12.6 gpm which wi.ll limit the calculated offsite doses to within 10 percent of 10 CFR 100 guidelines. Leakage in the intact loops is limited to 150 gpd. Zf the projected end of cycle distribution of crack indications esults 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-secondazy steam line break leakage to below 12.6 gpm.

PRESSURE BOUNDARY LEAKAGE indicative of of an impending gross any magnitude i.s unacceptable failure of the pressure since boundary.

it may be 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 pzessure 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 IDENTIFZED LEAKAGE and will be considered as a portion of the allowed lim'it.

3 4.4.7 CH=MZSTRY The limitations on Reactor Coolant System chemi.stry ensure that corzosion of the Reactor Coolant System is minimized and reduces the potent'al 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 li.fe of the plant. The associated effects of exceedi.ng the oxygen, chlor:de, 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 in"ervals without having a si.gnificant, effect. on the stzuctural 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 AHENDHENT NO 88g %44q 178

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ATTACHMENT 3 TO AEP:NRC:1166$

PROPOSED REVISED TECHNICAL SPECIFICATION PAGES

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EACTOR COOLANT SYST SURVEILLANCE RE UIREMENTS Cont nued

2. Tubes in those areas where experience has indicated poCential problems.

3. A tube inspection (pursuant Co Specif ication 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 ~ 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 roll 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 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.

e. Implementation of the steam generator tube/tube support plate interim plugging criteria for one fue1 cycle (Cycle 15) requires a 100% bobbin coil inspection for hot leg Cube 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~ox~ Ins ect on Results C-l Less than 5% 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 5S 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. 'OS, 454, 446, 477-,

EACTOR COOLANT SYSTEMS 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 R~e orts

a. Following each inservice inspection of steam generator tubes, there are any tubes requiring plugging or sleeving, the number of if tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

b. 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:

l. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged or sleeved.

C. 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 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).

e. The results of steam line break leakage analysis performed under T/S 4.4.5.4.a.l0 will be reported to the Commission prior to restart for Cycle 15.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-12 AMENDMENT NO. 403, 446, 477,

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CTOR C 0 SYS OPERATIONAL LEAKAGE IM TING CONDITION OR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEDGE, 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 15,

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.

~ACTI 0

a. 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 syst: em pressure isolation valve(s) leakage greater than the above limit, except when:

1. ,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 wi.th the leakage criteria.

Specification 3.4.6.2.e is applicable with average pressure wi.thin 20 psi of the nominal full pressure value.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-16 AMENDMENT NO. 462, 466, ~

Order dated Apri.l 20, 1981

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EAC OR COO SYST BASES 3 4 4 5 STEAM GENERATORS TUBE INTEGRITY 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 1. 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 oR 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 plant operation would be limited by the limitation of steam generator during tube leakage between the primary coolant system and the secondary coolant system. The allowabe primary-to-secondary leak rate is 150 gallons per day per steam generator for one fuel cycle (Cycle 15). Axial or circumferentia'ily 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. 'eakage 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 a'll tubes with imperfections exceeding the repair limit which is defined in Specification 4.0.5.4.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.lO.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-2a AMENDMENT NO. 403, 454, 446, ~

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REACTOR COOLANT SYSTEM BASES Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) for Fuel Cycle 15 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 will limit 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 fault'ed 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 indicative of of an impending gross any magnitude is unacceptable since failure of the pressure boundary.

it Should may be PRESSURE BOUNDARY LEDGE 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. Lealcage 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. 53, 444, ~

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