ML17333A819

Amend 215 to License DPR-58,revising TS to Allow Permanent Extension of Interim SG Tube Voltage Based Repair Criteria for SG Tubes Used in Cycles 13,14 & 15 at Plant

ML17333A819
Person / Time
Site:	Cook
Issue date:	03/13/1997
From:	John Hickman NRC (Affiliation Not Assigned)
To:
Shared Package
ML17333A820	List: ML17333A819 ML17333A821
References
NUDOCS 9703240209
Download: ML17333A819 (20)

Text

UNITED STATES

, NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2055&0001 I DIANA MICHIGAN POWER CO PANY DOCKET NO. 50-315 ONALD C COOK NUC AR PLANT UNI NO.

1 M ND E

TO FAC OP TING ICENSE Amendment No.

2" 5 License No. DPR-58 The Nuclear Regulatory Commission (the Commission) has found that:

A.

B.

C.

D.

E.

The application for amendment by Indiana Michigan Power Company (the licensee) dated June 19,

1996, and supplemented September 19,

1996, and December 20,

1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the'common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9703240209 9703i3 PDR ADCICK 05000315 P

PDR

2.

i Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.

DPR-58 is hereby amended to read as follows:

echnical S eci at'o s

3.

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 2i', are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

This license amendment is effective as of the date of issuance, with full implementation within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSION C~~

."John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

T ACHH NT TO CENSE AHE DHENT NO.

TO C

I Y OPE T

G IC NS NO DPR-58 OCKET NO. 50-315 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

jgHO~V 3/4 4-8 3/4 4-11 3/4 4-12 3/4 4-16 B 3/4 4-2a B 3/4 4-3

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>I4.4 REACTOR COOL STEM

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E (continued) 2.

Tubes in those areas where experience has indicated potential problems.

3.

h tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.. Ifany selected tube does not permit the passage ofthe eddy current probe for a tube inspection, this shall bc rccordcd and an adjacent mbe shaH be selected and subjected to a tube inslection.

4.

Tubes left in service as a result ofapplication ofthe tube support plate plugging criteria shaH bc inspected by bobbin coil probe during aH future refueling outages.

c, In addition to the sample required h 4.4.5.2.b.1 through 3, aH tubes which have had thc I." criteria applied wiH be inspected in the roll expanded region.

The roH expanded region of these tubes may be excluded &om the requirements of 4.4.5,2.b.l.

d.

The tubes selected as the second and third samples (ifrequired 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 ofthe tube sheet array where tubes with imperfections were previously found.

2.

The inspections include those portions of the tubes where imperfections were previously found.

c.

Implementation of the stcam generator tube/tube support plate plugging criteria requires a 100 percent 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 outside diameter stress corrosion crachng (ODSCC) indications, The dctertnination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.,

hspection of sleeves wiH foHow the hitial sample selection (1~ sample) and sample expansion requirements of Table 4.4-2.

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

ga~tc o

C-1 C-2 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

One or more tubes, but not more than 1% of thc total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% ofthe total tubes insiected are degraded tubes or more than 1%

of the inspected tubes are defective.

COOK NUCLEAR PLANT-UNIT1

'age 3/4 44 AMENDMENT-29S-, 21 5

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3/4 LIMITINGCOND S FOR OPERATION ANDSURVEILL REQUIREMENTS 3/4,4, REACTOR COOL, STEM VE LLAN E RE U

E ENTS (continued) 9,

~revju u rube is peruuuerr wilh rube suppers piers sieeves uud wilh rubesheer sieeves.

Tube support plate sleeves are centered about the tube support plate intersection.

Tubesheet sleeves start at the primaty Quid tubesheet face and extend to the free span region of tube above ihe tubesheet.

10.

be Su are is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. ht tube support plate mtersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:

Steam generator tubes, ~hose degradation is attributed to ouside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts willbe allowed to remain in service.

b.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2,0 volts willbe repaired or plugged, except as noted in 4.4.5.4.a.10.c below.

C.

Steam generator tubes with indications of potential degradation anributed to outside diameter stress corrosion cracking within the bounds ofthe tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair hmit may remain in service ifa rotating pancake coil or acceptable alternative inspection does not detect degradation.

Steam generator tubes, with indications ofoutside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair Hmit will be plugged or repaired.

Ifan unscheduled middle inspection is performed, the following mid~cle repair limits apply instead ofthe limits identified in 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c.

The upper voltage repair limit is calculated according to the methodology in Generic Lener 9545.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-11 AMENDhKNT~ 215

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3/4.4 REACTOR COOL YSTEM URV ILLAN E

E S (continued)

The mid~cie repair limits are detertnined from the following equations:

V

Vst, JttIJK CL - ar LO + NDE Gr CL CL-at Vsttat.

Vjttit- ( Vivat - Vsat, )

CL where:

VURL

~

upper voltage repair limit VLRL ~

lower voltage repair limit VMURL~ mid~cle upper voltage repair limitbased on time into cycle VMLRL~ mid~cle lower voltage repair limit bassed on VMURLand time into cycle ht

~

Length of time since last scheduled inspecuon during which VURLand VLRLwere implemented CL

~

cycle length (the time between two scheduled steam generator inspections)

VSL

~

structural limit voltage Gr

~

average growth rate per cycle length NDE

~ 95~rcent cumulative probabiliq allowance for nondestructive examination uncertainty (i,e., a value of 20-percent has been approved by NRC)2 implementation of these mid~cle repair limits should follow same approach as in TS 4.4.5.4.10.a, 4.4.5.4. 10.b, and 4.4.5.4.10.c.

The NDE is the value provided by the NRC in GL 9545.

COOK NUCLEAR PLANT-UNlT 1 Pace 3/4 4-11a NENDHF.NT 215

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aanutmLi t;VHD NS FOR OPERATION AND SUR

'/4 4 REACTOR COOL SYSTEM U V N

(continued) 12.

DDsmnce is thc distance &om the bottom of thc hardroll transition toward the bottom ofthe tubesheet that has been conservatively determined to be 1.11 inches (not including eddy current uncertainty).

g'I'ube is a.tube'with degradation, below the F~ distance, equal to or greater than 40%,

and not degraded (i.e., no indications of crackmg) within the F'istance.

13.

+~giair refers to sleevtng as described by'the reports listed in 4.4.5.4.c which are used to maintain a tube in service or return a tube to service.

Tubes with degradation.

indications of less than the plugging limit may bc preventively sleeved at the Owner's discretion.

This includes removal of plugs that were installed as a corrective or preventive measure.

h tube ittspection per 4.4.5.4.a.8 is required prior to returning previously plugged tubes to service. Further restrictions regarding identified indications and their proximity to the joint areas of various sleeving processes may be applicable.

b.

'Ihe 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-waH cracks) rettuired by Table 4.4-2.

C.

Steam generator tube repairs may be made in accordance with the methods described in either

%CAP-12623, %CAP-13088 (Rev. 3), or CEH-313-P.

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...3/4.4 REACTOR COOL SYSTEM U

L (continued) 4.4.5.5 jk~es a.

Following each inservioe inspection of steam generator tubes, ifthere are any tubes requiring plugging or sleeving, the~bcr 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 hscrvice inspection shall be hciudcd in the Annual Operating Report for the period in which this inspection was completed.

This rcport shall include:

1.

Number and extent of tubes and sleeves hspected.

2.

Locauon and percent ofwa114Cickness penetration for each indication ofan imperfection..

3.

Identification of tubes plugged or sleeved.

C.

Results of steam generator tube inspections which fall into Category C-3 and ratuire 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 ofthe tube degradation and corrective measures taken to prcvcnt rccurrcncc.

d.

Ifestimated leakage based on the projected endwf~cle (or ifnot practical, using the actual measured cndwf~cle) voltage distribution exceeds the leak limit(determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.

For implementation of the voltagekased repair criteria to tube support plate intersections, notify the Commission prior to returning the steam gcncmors to service should any of the following conditions arise:

1.

2.

If circumferential crack-like hdications are detected at the tube suppott plate tntcrscctlons.

3.

Ifindications are Identified that extend beyond the confmes of the tube support plate.

4.

Ifindications are identified at the tube support phte elevations that are attributable to primary water stress corrosion cracking.

S.

Ifthe calculated conditional burst probability based on the projected endwf~cle (or if Qot practical, using the actual measured endwf cycle) voltage distribution exceeds 1 x10, notify the Commission and provide an assessment of the safety significance of the occurrence.

COOK NUCLEAR PLANT-UiiIT1 Pape 3/4 4-12 NEND,"lENT~ 215

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~3/4.4 REACTOR COOL YSIZM D

0 R

P ON 3.4.6.2 b.

Reactor Coolant System leakage shall be Hmited to:

I'o PRESSURE BOUNDARYLEAKAGE, 1 GPM UNIDENTIFIEDLEhKAGE, C.

d.

e.

600 gallons per day total primatywmecondaty leakage through all steam generators and 150 gallons per day through any one steam generator, 10 GPM IDENTIFIEDLEAKAGEf'rom the Reactor Coolant System, Seal line resistance greater than or aqual to 2.27E-I ft/gpm'nd, The leakage from each Reactor Coolant System Pressure Isolation Valves specifie in Table 3.44 shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a Reactor Coolant System average pressure within 20 psi of the nominal full pressure value.

ao With any PRESSURE BOUNDARYLEPEAGE, be in at last HOT STANDBYwithin 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 BOUNDARYLEAI(AGE,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 STANDBYwithin. 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 />.

Ca With any reactor coolant system pressure isolation valve(s) leakage greater than the above limit, declare the leaking valve inoperable and isolate the high pressure portion of the affected system from the low prcssure portion by the use of a combination of at least two closed valves, one of which may be the OPERABLE check valve and the other a closed de~rgized motor operated valve; Verifythe isolated condition ofthe dosed dcmergized motor operated valve at last once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least HOT STANDBY within thc 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 />.

Specificatio 3.4.6.2.e is applicable with average pressure within 20 psi ofthe nominal fullprcssure value.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-16 hMENDi~iENT~,

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'4/4 jSAbM

'/4 4 REACTOR COOL SYSIZM 4.4.5 RA R

'Ihe Surveillance Requirctnents for inspection ofthe steam generator mbes ensure that the strucniral integrity ofthis portion of the RCS willbe maintained, 'Ihe program for inservicc inspection of stcam generator tubes is based on a modification of Reguiatoty Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the canditions of the tubes in the event that them is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inscrviee conditions that lead to corrosion.

Inservice inspection of stcam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that conective measures can be taken, The plant is expected to be oiieratcd in a manner such that the secondary coolant willbe maintained within those chemistry limits found to tesult ln negHgible cortosion of the steam generator tubes. Ifthc secondgy coolant ciiemistty is not maintained within these parameter limits, IocaHzed corrosion may Hkely result m sttess conosion cracking. 'Ihe extent ofcracking during plant operation would be limited by the limitationofsteam generator tube leakage between the primary coolant system and the secondaty coolant system. The aHowable primaty<mecondary leak rate is 150 gaHons per day per steam generator.

Axial or circttmferetttiaHy.oriented cracks having a primary-to-secondary leakage less than this limitduring operation wiH have an adequate margin of 'safety to withstand the loads imposed during normal operation and by postulated accidents.

Leakage in excess of this limitwillrequire plant shutdown and an inspection, during which the leaking tubes willbe 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 reinsuuement of auxiliary or main feed low control and steam control.

Wastage-type defects are unlikely with the aH volatHe torment (AVT)of secondary coolant.

However, even if a defect ofsimilar type should develop in service, itwillbe found during scheduled inservice steam generator tube examinations.

Plugging or sleeving wiH bc required for aH tubes with imperfections exceeding the tepair limit which is defined in Specification 4.4.5.4.a.

Steam generator tube inspections ofoperating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube waH thickness.

COOK NUCLEAR PLANT-UNITI Page B 3/4 4-2a AMENDMEhT-8$; 215

, 3/4 BASES 3/4.4 REACTOR COOL

$ySTEM

/44 E MG E

T t'ed The voltage-based repair limits of these surveillance requirements (SR) implement the guidance in GL 9545 and are'applicable only to Westinghousedesigned steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections.

The voltagekased repair limits are not applicable to other forms of SG tube degradation nor are they <<pplicable to QDSCC that occurs at other locations within the SG. Additionally, the repair criteria apply only to hdications where the degradauon mechanism is dominantly axial ODSCC with no NDE detectable cracks extending outside the thickness of the support plate.

Refer to GL 9545 for additional description of the degradation morphology.

hnplementation of these SRs requires a derivation of the voltage structural limit from the burst versus voltage empirica correlation and then the subsequent derivation of the voltage repair limitfrom thextructural limit(which is then implemented by this sutveillance).

The voltage structural limit is the voltage from the burst pressure/bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95~t tolerance bound for tubing material properties at 650'F (i.e., the 95-percent LTLcurve). The voltage structural limitmust be adjusted downward to account for potential degradation growth during an operating interval and to account for NDE uncertainty.

The upper voltage repair limit;VUltL,is determined from the structural voltage limitby applying the followingequation:

E VURL VSL - VGr - VNDE where Vo, represents the allowance for degradauon growth between inspections and VNDErepresents the allowance for potential sources oferror in the measurement ofthe bobbin coil voltage. Further discussion ofthe assumptions necessary to determine the voltage repair limit are discussed in GL 9M5.

The mid~cle equation in SR 4.4.5.4.10.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.

SR 4.4.5.5 implements several reporting requirements recommended by GL 9545 for situations which the NRC

'wants to be notified prior to returning'he SGs to service, For the purposes ofthis reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected endwf~cle (EOC) voltage distribution (refer to GL9545 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to rettttning the SGs to service.

Note that ifleakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.l and 6.a,3 reporting criteria, then the results of the projected EOC voltage distribution should provided per the GL section 6.b (c) criteria.

%des experiencing outer diameter stress corrosIon cracking within the thickness of the mbe support plates are plugged or repaired by the criteria of 4.43.4.a. 10.

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, 3/4 BASES 3/4.4 REACTOR COOL SYSTEM

/4.4 AM ed j

Whenever the results of any stcam generator tubing inservicc inspection fail into Category C-3, these, results will be promptly reported to the Commission pursuant to SpecUication 6.9.1 priorto resumption ofplant operation. Such cases willbe considered by the Commission on a case4y~

basis and may result in a requirement for analysis, laboratory examinations, tats, additional eddywurrent insIcction, and revision of the Tcchnical Specifications, if scccssaty.

Degraded steam generator tubes may be repaired by the hstallation ofsleeves which span the section of degraded stcam generator tubing. h steam generator tube with a sleeve installed tnccts the structural rcquumncnts of tubes which are not degraded, To determine the basis for the sleeve plugging limit, the minimum sleeve wall thickness was calculated in accordance with Draf't Regulatory Guide 1.121 (August 1976). In addition, a combined allowance of 20 percent of wall thickness is assumed for eddy current testing inaccuracies and continued operational degradation per Draft Regulatory Guide 1.121 (August 1976).

'Ihc following sleeve designs have been found acceptable by the NRC staff:

1.

Westinghouse Mechanical Sleeves (WCAP-12623) 2.

Combustion Engineering Leak Tight Sleeves (CEN-313-P) 3.

Westinghouse Laser Welded Sleeves (WCAP-13088, Rev. 3)

Descriptions of other future sleeve designs shall be submitted to the NRC for review and approval in accordance with 10 CFR 50.90 prior to their usc in the repair of degraded stcam generator tubes.

The submittals related to other sleeve designs shall be made at least 90 days prior to use.

COOK NUCLEAR PLANT-UNIT1 Page B 3/4 4-2C AMENDMENT 2)5

3/4.4 REACTOR COOL YSTEM

/4.4

/4.4.6 E

G The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundaty.

Qesc detection systems are consistent with the recommendations of Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Daeaion Systems, May 1973.

/4 0

Indusay experience has shown that whBe a Hmited amount of leakage is cxpeaed from the RCS, the unidentified portion of this leakage can be reduced to a threshold value ofless than I gpm. Thh threshold value is sufficicntly Iow to ensure early deteaion of additional leakage.

The 10 GPM IDENTIFIEDLFM(AGEHmitations provides allowance for a Hmited amount ofleakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage

'daection systems.

l The Hmitation on seal line resistance ensures that the seal line resistance is greater than or equal to the resistance asst tmed in the minimum safeguards LOCAanalysis. This analysis assumes that all ofthe flowthat is diverted from the boron injection Hne to the seal injection line is unavailable for core cooling.

Maintaining an operating leakage limitof 150 gpd per stcam generator (600 gpd total) willmittimixc the potential for a large leakage event during steam line break under LOCAconditions.

Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate followinga steam line rupture is limited to below 8.4 gpm which wiH ensure the calculated offsite doses willremain within 10 percent ofthe 10 CFR 100 requirements and that control zoom habitability continues to meet GDC-19. Leakage in the intact loops is limited to 150 gpd. Ifthe projected end of cycle distribution of crack indications results in primary-to-secondary leakage greater than 8.4 gpm in the faulted loop during a postulated steam line brcak event, additional tubes must be removed from service in order to reduce the postulated primary~secondar stcam line break leakage to below 8.4 gpm.

Also, the 150 gpd leakage limitincorporated into this specification is more restrictiv than the stanthrd operating leakage Hmit and is intended to provide an additional margin to accommothte a crack which might grow at a greater than expected rate or unexpectyHy extend ouside the thickness of the the tube support plate.

Hence, the reduced Ieakage limit, when combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced In service, it wiH be detected and the plant shut down in a timely manner.

PRESSURE BOUNDARYLEAIChGEofany magnitude is unacceptable since it may be indicative ofan impending gross failure ofthe pressure boundary.

Should PRESSURE BOUNDARYLEAI&GEoccur through a component which can be isolated from thc 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.

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