ML17329A422

Proposed Tech Specs Re Plugging Criteria for SG Tubes

ML17329A422
Person / Time
Site:	Cook
Issue date:	03/20/1992
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17329A421	List: ML17329A420 ML17329A422 ML17329A423 ML17334B415
References
NUDOCS 9203260130
Download: ML17329A422 (32)

Text

ATTACHMENT 2 to AEP:NRC:1166 PROPOSED REVISED TECHNICAL SPECIFICATIONS PAGES 9203260130 920320 PDR ADOCK 050003l5 PDR

L I G CONDITION FO OPERAT 0 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3 and 4.*

ACTION:

With one or more steam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing T,~ above 200 F.

0 SURVEILLANCE RE UIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirement of Specification 4.0'.

4.4.5.1 Steam Generator Sam le Selection a

Ins ectio

- Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sam le Selectio and Ins ectio

- The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.

The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.

The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators'the tubes selected for these inspections shall be selected on a random basis except:

I a.

Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

b.

The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

1.

All tubes that previously had detectable wall penetrations (greater than or equal to 20%) that have not been plugged or repaired by sleeving in the affected area.

• This Specification does not apply in Mode 4 while performing crevice flushing as long as Limiting Conditions for Operation for Specification 3.4.1.3 are maintained.

COOK NUCLEAR PLANT UNIT 1 3/4 4-7 AMENDMENT NO.

CO CO SYS SURVEILLANCE RE UIREMENTS Continu d 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,'his 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:

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.

Implementation of the steam generator tube/tube support plate alternate plugging criteria requires a 100$ bobbin probe 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.

An inspection using the rotating pancake coil (RPC) probe is required for all tube support plate intersection indications for bobbin coil signal amplitude between 1.5 and 4.0 volts provided the tube is to be left in service.

The RPC results are to be evaluated to establish that the principal indications can be characterized as ODSCC.

Once an indication is characterized as

ODSCC, RPC inspections will be performed on this indication at alternate subsequent refueling outages.

For tubes that will be administratively plugged, no RPC inspection is required.

The results of each sample inspection shall be classified into one of the following three categories:

~Cate ~ory Ins ection esults C-1 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 5% and 10%

of the total tubes inspected are degraded tubes.

C-3 AMENDMENT NO. QS, 444 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

S o

u Note:

In all inspections, previously degraded tubes must exhibit significant (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Ins ection Fre uencies

- The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.

Subsequent inservice inspections shall be performed at intervals of not less than 12 or more than 24 calendar months after the previous inspection.

If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-l category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

b.

If the results of inservice inspection of a steam generator conducted in accordance with Table 4'-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months, Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1.

Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.

2.

A seismic occurrence greater than the Operating Basis Earthquake.

3.

A loss-of-coolant accident requiring actuation of the engineered safeguards.

4.

A main steam line or feedwater line break.

d.

Tubes left in service as a result of application of the tube support plate alternate plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-9 AMENDMENT NO.

Q8

S ST S

V C

R IREMENTS Co t nued 4.4.5.4 a ce Crite a

~

As used in this Specification:

contour of a tube or sleeve from that required by fabrication drawings or specifications'ddy-current testing indications below 20% of the nominal wall thickness, if detectable, may be considered as imperfections.

2.

general corrosion occurring on either inside or outside of a tube or sleeve.

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 wall thickness affected or removed by degradation.

5.

Defect means an imperfection of such severity that it exceeds the repair limit.

Re air Plu in L mit 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.

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.10 for the plugging limit for use within the thickness of the tube support plate.

7.

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. 08, 444

C t u

around the U-bend to the top support of the cold leg.

For a

,tube in which the tube support plate elevation alternate 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.

~sleevln 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.

u 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 alternate 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 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 4.0 volts, 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 120 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.

2.

A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 4.0 volts.

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.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-11 AMENDMENT NO. 98,

REACTOR-COO SYS S

V L

IR 4.4.5.5

/eeoc a e 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.

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:

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 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 alternate 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).

COOK NUCLEAR PLANT - UNIT 1 3/4 4-12 AMENDMENT NO. 88, 444

TABLE 4 4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No. of Steam Generators per Unit First Inservice Inspection Second

& Subsequent Inservice Inspections Yes Four Two One~

Table Notation:

1.

The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3

N%

of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.

Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators.

Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.

2.

The third and fourth steam generators not inspected during the first inservice inspection shall be inspected during the second and third inspections, respectively.

The fourth and subsequent inspections shall follow the instructions described in 1 above.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-13 AMENDMENT NO. B&

1ST S

P I SPECTION TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTIO 2ND SAMPLE INSPECTION 3RD S

PLE INSPECTION Sample Size A

minimum of S

Tubes per S.G.

Result C-1 C-2 C-3 Action Required None Plug or sleeve defective tubes and inspect additional 2S tubes in this S.G.

Inspect all tubes in this S.G., plug or sleeve defective

tubes, and inspect 2S tubes in each other S.G.

Result N/A C-1 C-2 C-3 All other S.G.s are C-1 Action Required N/A None Plug or sleeve defective tubes and inspect additional 4S tubes in this S.G.

Perform action for C-3 result of first sample

,None Result N/A N/A C-1 C-2 C-3 N/A N/A Action Required N/A N/A None Plug or sleeve defective tubes Perform action for C-3 result of first sample N/A N/A Prompt noti-fication to NRC pursuant to specification 6.9.1 Some S.G.s C-2 but no additional S.G.

are C-3 Additional S.G

~ is C-3 Perform action for C-2 result of second sample Inspect all tubes in each S.G.

and plug or sleeve defective tubes.

Prompt notific-ation to NRC pursuant to specification 6.9.1 N/A N/A N/A N/A n%

ere s t e num er o steam generators n t e un t, an n

s t e num er o

steam generators inspected during an inspection COOK NUCLEAR PLANT - UNIT 1 3/4 4-14 AMENDMENT NO.

CT CO ST 4 4 C

COOLANT SYSTEM LEAKAG LG DC 0SSEMS LIMITING CONDITION FOR OPERATION 3.4.6 '

The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a

~

One of the containment atmosphere particulate radioactivity monitoring channels (ERS-1301 or ERS-1401),

b.

The containment sump level and flow monitoring system, and c

~

Either the containment humidity monitor or one of the containment atmosphere gaseous radioactivity monitoring channels (ERS-1305 or ERS-1405).

APPLIC BILI MODES 1, 2,

3 and 4.

ACTION:

With only two of the above required leakage detection systems

OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous and/or particulate radioactivity monitoring channels are inoperable; otherwise, 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 />.

SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

a

~

Containment atmosphere particulate and gaseous (if being used) monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, b.

Containment sump level and flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months, C.

Containment humidity monitor (if being used)

- performance of CHANNEL CALIBRATION at least once per 18 months.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-15 AMENDMENT NO. 400,

EAC COO SYSTEM 0 ERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION I

3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, cd 600 gallons per day total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator, d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.

52 GPM CONTROLLED LEAKAGE.

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:

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 />.

With any reactor coolant system 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 with the leakage criteria.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-16 AMENDMENT NO.

EAC OR COOLANT SYSTEM 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 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 allowable primary-to-secondary leak rate is 150 gallons per day per steam generator.

Axial or circumferentially oriented cracks having a primary-to-secondary leakage less than this limit during operation will have anadequate 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.

REACT CO SYST

~SEES 3 4 4 5

S EAM GENERATORS TUBE INTEGRITY (Continued)

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations,

tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-2b AMENDMENT NO.

CTO C

SYS BAS S

Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) 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 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 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.

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 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

0 0

Q~ES The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3 4 4 8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters of the Cook Nuclear Plant

site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.

Offsite doses following a main steam line break are limited to 10 percent of the 10 CFR 100 guideline.

This restriction is based on the results of a Donald C, Cook Nuclear Plant site specific radiological evaluation that assumes a primary coolant iodine activity level corresponding to 1 percent fuel defects (approximately 4.6 microCurie/gram DOSE EQUIVALENT I-131),

rather than a specific activity of 1.0 microCurie/gram DOSE EQUIVALENT, and a post-accident primary-to-secondary leak rate of 120 GPM in the faulted loop.

Reducing T>>9 to less than 500 F prevents the release of activity should a

0 steam generator tube rupture since the saturation pressure of the primary coolant is below the liftpressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive'pecific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-5 AMENDMENT NO.

ATTACHMENT 3 to AEP:NRC:1166 CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES

REACTOR COOLANT SYSTEM SHVIEILLAiCE RE UIRENENTS Continued l.

All tubes that previously had detectable wall penetrations (greater than or equal to 20%) that have not been plugged or repaired by sleeving in the affected area.

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.

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 par"ial 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.

~Q QV 2.

The inspections include those portions of the tubes where f4+A~y~

i imper fee tions were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

~Cate orv Ins ection Results C-1 Less than St 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 1'0 of the inspected tubes are defective, Note:

In all inspections, previously degraded tubes must exhibi" significant, (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-8 AMZuDMEVr No.

gg,.~~t

'I

NElPii)R Cggt~yy Sy~

~gyPt ~g ~<~J~~~g~S

(~"n-tnuo 4.4.$.3 fnsnect.on F e uencies

>i. e above reouiref inservice insyec: ons Of'%am gene.

c r tuoes snai l be Oert'OP4td at M4 tollwinq t'requenctes:

De firSt fnSarviCe inS:e.tiOn sha11 be Pert'O~ at.e.

5 fffectfve Full roe.

Nonths but ~f thfn 24 calencar aenths o~

inftfal critical i:y.

Subsequent inservfce insoectfons shall be per. ~ at intervals ot'ot less than 12 nor cere than 24 calencar nenths after the prevfous insoec fon.

lf' consecu-fve inspections t'o!1o&nq service under AYT con4ftfons, not incluCfnq the Omsemica insyec:fon, reSult in 411 insoec:fon resu(ts failing int" the C-1 category or if' cons~cfve insoectfons Ceans+"aa that previously observed Cegradatfon los not contfnueC and no acdftfona1 d~radatfon has oc~eC,

~

.e insoectfon interval may be extenctt4 U a aaxfmun ot'nce

)er cQ mntns.

C.

t..e reSultS C.:he inSerV<Ce nSCeC.fOn Of 4 SWr~

generator corcuc.ec in ac-orzrce ri th Tab!e 4.'

at 40 nanth

<ntarval5 fali in ~tegory C-3, the inspectfon t'rtquency sr'!1 i"e fr ~. eascc

" 4 least once vcr 29 4nths.

iite tnc. Case lrt ihS:cotton freouenCy sral'i 4py! y untf1 the suesequent,.insoec

.Ons tatfsfy the Crt teria

.O.

Spect...'catfon 4.4.5.3.4; the i'n:erval may then be extenct to 4 mx.mum ot'nce;er 4~ months.

Acdftfonal, unschaCul& inse~fce inspec.fons sna11 be per.or.ec oh each 5 east generator in accordance ef th the t'irs saaple fhs ec:fon spe f~fec in Ta'hie 4.4 2 cur fag the scutum suosecuen:

to any ot'he i'olleefnq conoftfons:

--- =."""7>"--h iaaf-to.-.secondary.Oe.l eag. Jot; includfng 1eaks orfqfnatfng t'ra woe-Ca-tuoe sheet +14'i) in cicadas't' the 1 iafts ot'pecfffcatfon 3.4.5.2.

A 5ef safe oc:~rance greeter than ta Oyeritfng Sasis Earthquake.

A losswt'moolah accfdent ctqufrfncl acestfon of'"e engfneereo sai'equals.

4, A main stean line or 'eedmwr 1ine break.

d.

Tubes left in service as a result of application of the tube support plate alternate plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.

4 e

Aaaachmnc, So

REACTOR COOVQfT SvST~q Sl;RVEILLAfCE REOUZREv~~iS Continued 4,4, 5 4 Acceocance Cri aria hs used

.'n this Specif'cation:

I oerfeccion means an exception to the dimensions, finish or contour of a cube or sleeve from that required by fabrication dravings or specificacions.

Eddy-current casting indicacions belov 20i of che nominal vali thickness. if dacectabl ~,

may be considered as imperfections.

2.

Decradacion means a serviceeinduced cracking, vastage, vea.

o.

general corrosion occurring on either inside or outside of a tube or sleeve.

Dexradcd Tube o.

Sleeve means an imperfection greater than or equal co 204 of the nominal vali chickness caused by de grada cion.

Percent Degradation means che amount of the vali thickness affected or removed by degradation.

~

~

~

l

~

te!eo= teats an Leper!sation e! auth severity that it exoeeas the rape-ii'n.it.

e Re ai /Plu in Limit means che imperfection depth at or beyond vhich the tube or sleeved cube shall be repaired or removed from service.

Any tube vhich, upon inspection, exhibits cube vali degradation of 40 percent or more of che nominal cube vali chickness shall be plugged or repaired prior to recurning the steam generacor to service.

any sleeve vhich, upon inspection, exhibits vali degradation of 29 percent or more of the nominal vali thickness shall be ylugged prior co returning the sceam generator co service.

In

addition, any sleeve exhibiting any measurable vali loss in sleeve expansion transicion or veld cones shall b>> plugged.

9.

~tlat<<n a toke is peen!stat only in areas there tha sleeve, syans che cubesheet area and vhoae lover ]oint is at, the primary fluid tubesheet face.

COOK NUCLMt PLY - UNIT 1

~i4 4-10 ANBCMBIT 80. 98.

f~~t

,::elBjo

'i" ic. leaks or contains a defect large enough co affect its structural integrity in the event of an Operating Sasis

":...,. '.. -.:,.'.~.Larzhquaka; a 4oaa of c'co%'am "acct'dane,'t 5'r"i a'tesi 'line or J

!aeavataz line break as speoi!iea in 4.a.5.3.o, shove.

( r,!p tube or sleeve froa the point of entry (hot leg aide) complecely around the U-bend to the top support of the cold leg.

tNSERT A

d.

Implementation of the steam generator tube/tube support plate'lternate plugging criteria requires a

100% bobbin probe 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.

An inspection using the rotating pancake coil (RPC) probe is required for all tube support plate intersection indications for bobbin coil signal amplitude between 1.5 and 4.0 volts provided the tube is to be left in service.

The RPC results are to be evaluated to establish that the principal indications can be characterized as ODSCC.

Once an indication is characterized as

ODSCC, RPC inspections will be performed on this indication at alternate subsequent refueling outages.

, For tubes that will be administratively plugged, no RPC inspection is required.

INSERT 8 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.10 for the plugging limit for use within the thickness of the tube support plate.

INSERT C

For a

tube in which the tube support plate elevation alternate 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.

~R COOLLY gyS~S CX R QTR~S Conc tnued

~e steam generator shall be determined OP~LE afcet complete che

~

nC che corresponding actions (plugging or sleeving all cubes ezceedi

~

repair limit and all cubes containing through-vali cracks) tnt

<<9u<

ed by Tabl ~ 4.4 ~ 2.

Sceaa generator c be repairs aay be aada in accordance vich che methods described in eicher VCR' 12523 ot C&-313-P.

<.4.5.f R ~'00l'll ao Folloving each

'.nservica Insyeccion of scca! jeneracor

cubes, iaaf ~ a'r ~ any cub ~ s regui nf pluging or sleeving, the number of CQes plugged ot sleeved in each scaca jeneracot shall be reported co he Cocaission vichin 19 days.

~ e cocy I. e ro iu s

Of che s cease gene racot cube Saservice inspec 'n 5ha~ 'e

'nc~ uded in che equal Operating Report period in vn.'cn

..".'s inspection vas comylecad.

This tepott, inc 'da:

fot.ie shall Nuaber and extent of cubes inspected.

h Loca..'an and percenc of vali ~ thickness penectacioa fot ind'cation of an imperfection.

C 4

L eaA c

~

iden i.".cacion of cubes pluged ot sleeved.

~

Resul s o.

scaaa generator cube insyec.'oas vhich fall into Category C.3 and require ptoayc nocificacioa of the Coaaissioa shall be reported pursuant co Specificatioa 6.9.1 ptiot Co reauapcion of a de plant operacioa.

The vtittaa follovuy of this report shall ovida sctiptioa of invescigatioas conchcted to dataaiaa cause of the 5

pt cube degtadacioa and cotteccivo meaautea ca3caa co ptaveat tecutteace.

d The results of inspections performed under 4.4.5.2 for all tubes in which the tube support plate alternate 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).

C~" ~4LFm rLagt - mZT l 3/4 4 ll gg,,f5 '

e ube Su o t a

e te u

C 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 alternate 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 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.

l.

A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 4.0 volts, 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 120 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.

2.

A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 4.0 volts.

REACTOR COOLANT SYSTEM PERATI LEAKAGE LIMITING CONDITION FOR OPERATION 3,4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNCARY LEAKAGE, b,

C.

d.

1 GPH UN IDEN T I F I ~5 LEAKAGE, oo to al arista ry-to-secondary leakage through all steam gener-ators and %%gallons per day through any one steam generator, i'/0 10 GPM IOENTIFIEO LEAKAGE from the Reactor Coolant System, and e.

52 GPH CONTROLLED LEAKAGE.

f.

1 GPM leakage

.'rom any reactor coolant system pressure isolation valve speciried in Table 3.4-0.

APPLICABILITY:,%0ES !, 2, 3 and 4

ACTION:

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 COLO 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

,reater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to wfthfn limits wfthfn 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) leak-age 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

>> o satisfy ARA 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 shying that the method is capaole of demonstrating valve compliance with the leakage criteria.

0.

C.

COOK - UNIT 1

3/4 4-16 8+M goasOJ'y 4,

b)

REACTOR COOLAHT S'EST~

USES

4. 4 5 STEP&

C fERA QRS

~~

5E j.c fECRI '

The Suveil'.anca Requirements for inspection of the steam generator cubes ensu e chat C.".e scr ctursL i,ntsgricy of this portion of the RCS viLL be maintained.

The program for inservics inspection of steam generator cubes is based on a modification of ReguLatory Cuide L.d3, Revision L.

Inservice inspection of steam gereracor cubing is essential in order co maintain surveiLLance of che cordic'ons of the tub4s in the event that share is evidence of mechanical damage or progress jve degradation due to design, manufacturing errors, or inservice conditions c&t load to corrosion:

Inservice inspection of steam generator cubing also provides a aaans of characterize.ng the nature and cause of any cube degradation so that corrective measures can be taken.

.he plant is expected

o be operated in a manner such Chat the second.

ary coolanc vill be ma;.".calmed vichin chose chemistry Limits found Co resul.

in negL'gibl ~ corros'oon of che steam generator cubes.

If the secondary (0

cooisnc ohscisep is

..o saincsinad v'ic lin chas ~ possessor iini'cs, locsiisod corrosion may Likely

". ~suit in scresi corrosion cracking.

Tha extent, of s'P cracking during Piac:

oP grecian vouid bo linseed by Cha 1inieaeion at scssc gona sco.r cuba leaks

~ bacvsan ha priss coaisne s scan snd cha secondary '

coolanC s

O'Cem oj Co ~ se dary a

1 ype sc am gene

).

r sav p

.m

- o-da a

e 1

s an is im d

d p

te gi of saf d

d o

pe a

b s

la d

ci nt rati ala ha e

n ce c 'c

.se da a

o 5

al er hy r

st e

Co r+

et ia on oni ors f eam or ex 0

5 vi re ir pl u

o d a6 un e

d ect n,

1 e

nd h 'steam generator vhile undergoing crevice flushing in Hoda 4 is available for decay heat removal and is operable/operac'.."g upon reinstatement of auxiliary or main feed flov control and scaafs control.

Ascage-cypa defects are unlikely vith cha all volatile treatment (AVT) ot secondary coalane.

Hovavar, evan it a dataee ot

~iniiar cypa should develop in service, ic vill ba tound during scheduled Insaryics scads gsnoracoc cubo...

'exam'.nations.

Plugg'ing or slieving vill be rejuired for all tubes vith imper ~

~ ~

factions exceeding cha rtpair L4%ic vhich is d4fin4d in Specif icacion

.....,4..4.>.4..ua.:c,.:)~~. )curator.. Cube AsoyecsimnS deaf aParasiagn.P4ncs.

han..-

demonscraced the capability to reliably decect degradation chat has penecraced 20% 'of che original t&e vali thickness.

t'henever the results of any sceaa generator tubing inservfce inspection fall inco Caragory C ~ 3, these results vill ba promptly reported co the Commission pursuanc co Specification 6.9.1 prior ro resumption of plant operacion.

Such cases vill be considered by cha Commission on a case-by. case basis and may result in a requirement for analysis, laboratory exalinacions tests.

additional eddy-current inspection, and revision of tha Technical Specifications, if necessary.

COOK meme, sian

- mIT 1 he%~

Ho. 51)p St

The allowable primary-to-secondary leak rate is l'50 gallons per day per steam generator.

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.

anSA'7 F

diameter stress corrosion cracking within the Tubes experiencing outer amete thickness of the tube support plates are plugged or repaire y t e cri e of 4.4.5.4.a.10.

Q\\+

R~N CMLAÃl'YST94 e to sta ener$ t

-ube akage li of L QPH a11 aalu ge ator not fs atad f" e the ensure t the d

age c:

ibutfon

~ tug eaka

~ill."lief' to a

s fraction Part 0 Ifef n

'event f

haft.

a s

gener or Me pture ot

.aan lf break.

lfe fs c-sistan >f4'.

ass4act ns used f

the an sfs of ese a

<ga d leak e lfmf per sta generator nsures at sta gener r

fs ntafne in the nt of a n sta line ure o

under fons.

dents.

tu CA PRasSURK scu~OARY I.=-~mac of any eagnftude fs unacceptable since it any'e fndfcatfve of an ioccndfng gross fafIure of the pressure

boundary, Should PRESSURE 8QUHOARY L=AVGF. occur through a coaponent vhfch can be fsolatad froa the balance of the Reactor Coolant Systaa, plant, operatfon may continue provfced the 1eakfhg c cgonent i s procptly fsol atad freya the Reactor Coolant Systaa since fsolatfon ranoves

",e source.of potantfal failure.

M C

~

The Survef11anca RarufranenM for RQ Pressure lsolatfon Valves provide added assurance of valve intagrity thereby reducing the pMabflfty of gross valve faflure and consaouent fntarsystaa I.XA.

Laakage f~ the RCS Pressure i

Isolatfon Valves ls tOE!fTTFi:-9 I.~QE and ~f 11 be considered as a portion of the all+red 1fait.

3/4. 4. 7 CH~ISTAY The 1fafWtfons on Reactor Coolant Systaa cheafstry ensure that corrosion of the Reactor Coolant Systa! fs efnfafzed and reduces the potantfal for Reac-tor Coolant Systaa leakage or failure due to stress corrosfon.

Hafntafnfng the cheafstry ~tthfn the Steady State LfafQ provfdes adequate corrosion pro-tactfon to ensure the structural fntegrfty of the Reactor Coolant Systole over the 1ffa of the plant.

The 4554Cfated effects of excaedfng the oxygen, chloride, and fluoride 1fafts are tfae and teayerature dependent.

Corrosion studfes she that operatfon cay be contfnuect Wth contalfnant concantratfon levels fn excess

- of'he Shee}y State Lfafts,"uy'o: Me 7ranafent l.faith,.',or the s~i fact l,faitaa tfae fnta&als without having a significant effect on the structural fntagrity of the Re~t Coolant System.

The tfee interval permit ing continued operation

" 'v4'thorn A"mtv+fons"of'Ne.'7resfet."L+ci"provides"4foe.for" 0akQg'correc- '

tive ac:fons Co restore the contalfnant concantrations to vfthfn 10 Staacy 5Cata Liei&.

The surveillance recIugregents provide adequate assurance that concentrations in excess of the Liagts viLL be detected in sufficient tiae to take cor ective ac ion.

3/4. 4. 8 SPIC'..=TC AC ~.'1Z~g

~ L&itations on the speci c ac vity of the priaary cooLant, ~nsure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> do~es at:he site boundary viLL not exceed an appropri4teLy snaLL f:action of Par:

'.00 Limits foLLoving a stean genera(or tube

.capture acc'dent Ln con]unction fait'h an aswned steady state priaag-to-secondary steca generator Leakage rate of l.0 CPA.

The values for the L'a'=s on speci" c activ'-ty represent interest Liaf.ts based upon 4

paraae cr c evaluation by the NRC of r~ ical site l,ocar.'ns These values ar conse rvac ve in tha 5 pec c

5 i I paraae cars of the Cook Huc Lear PLant s i:e, such as si=e bou.".da.-y location and aeteorological conditions, vere not considered in zh's evaluation.

The

%C is f'nali in'ite specific critar'4 which viLL be used as

.he basis or he reevaLuation of the specific activity Limits of this si: ~,

This raev4luat on a4y result, in higher Lm Reducing T

=o '.ass

".han 500 F prevents the r ~ lease of act'vicy should 4

5 teach gene 1

o c

' "e

'p "

e 5 ince he saturation pressur ~ of the priaary cooLan is belov:..e pressure of he auospherfc stem relief val.ves.

The survef,lienee req ireaents provide adequate assurance that excessive speci 'c act'vi y Lave.'s in:h~ priaary coolant viLL be detec ad in suf c ent t ae to iice cor ective act'on.

Lnforaaclon obtained on iod.'.-.e sp'.king viLL be sad to.assess the parameters associated 4th spiking phenoaena.

A reduction

.'n frequency of isotopic analyses foLLoving pave".

changes may be peraissible if Justi 'ed by the data obtained.

'OCK HUCLZAR PLAÃr - UNIT I.

g 3/4 4-5 Amendmene Ne. f/f

Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) will minimize the potential~ a large leakage event during steam line break under LOCA conditions.

Eased o~n 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 t

b 1 w 120 gpm in the faulted loop and 150 gpd per steam generator in the te10 intact loops, which will limit offsite doses to within 10 percent of the CFR 100 guidelines.

If the pro)ected end of cycle distribution of crack indications results in primary-to-secondary leakage great'er 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.

E Offsite doses following a main steam line break are limited to 10 percent of the 10 CFR 100 guideline.

This restriction is based on the results of a Donald C.

Cook Nuclear Plant site specific radiological evaluation that assumes a primary coolant iodine activity level corresponding to 1 percent fuel defects (approximately 4.6 microCurie/gram DOSE EQUIVALENT I-131),

rather than a specific activity of 1,0 microCurie/gram DOSE EQUIVALENT, and a post-accident primary-to<<secondary leak rate of 120 CPM in the faulted loop.

ATTACHMENT 4 to AEP:NRC:1166 VCAP-13187, "D. C.

COOK UNIT 1 STEAM GENERATOR TUBE PLUGGING CRITERIA FOR INDICATIONS AT TUBE SUPPORT PLATES" (PROPRIETARY)

VCAP-13188, "D. C.

COOK UNIT 1 STEAM GENERATOR TUBE PLUGGING CRITERIA FOR INDICATIONS AT TUBE SUPPORT PLATES" (NON-PROPRIETARY)

ATTACHMENT 5 to AEP:NRC:1166 WESTINGHOUSE AUTHORIZATION LETTER CAV-92-282 AND ACCOMPANYING AFFIDAVIT, PROPRIETARY INFORMATION NOTICE, AND COPYRIGHT NOTICE