ML20077C570

Proposed Tech Specs Re Renewal of SG Tube Interim Plugging Criterim for Unit 1 Cycle 9

ML20077C570
Person / Time
Site:	Catawba
Issue date:	11/29/1994
From:	DUKE POWER CO.
To:
Shared Package
ML20077C565	List: ML20077C561 ML20077C570 ML20077C577
References
NUDOCS 9412050228
Download: ML20077C570 (27)

Text

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U. S. Nuclear Regulatory Commission Page 4 D. L. Rehn, being duly sworn, states that he is Site Vice-President, Catawba Nuclear Station; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission this revision to the Catawba Nuclear Station Technical Specifications, Appendix A to License Nos. NPF-35 and NPF-52; and that all statements and matters set forth therein are tme and correct to the best of his knowledge.

'D. L. Rehn i

Subscribed and sworn to before me this 29Aday of M k__,1994.

t Ohirdn k.

Notary PuNic My Commission Expires:

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9412050228 941129 DR ADOCK 05000413 PDR

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ATTACIIMENT 1 i

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LIMITING CONDITIONS FOR OPERATION AND SURVElllANCE REOUTREMENTS i

j SECTION PA_qE A

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1/4.4.2 SAFETY VALVES l

Shutdown................................................ 3/4 4-7 l

0perating............................................... 3/4 4-8 l

3/4.4.3 PRESSURIZER............................................. 3/4 4-9 3/4.4.4 RELIEF VALVES.......................................... 3/4 4-10 l

3/4.4.5 STEAM GENERATORS........................................ 3/4 4-12 l

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-17 i}

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION..................... 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................... 3/4 4-19.

j Ope rati onal Le akage..................................... 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.... 3/4 4-22 3/4.4.7 CHEMISTRY............................................... 3/4 4.-24 I

TABLE-3.4-2 REACTOR COOLANT SYSTEM' CHEMISTRY LIMITS............. 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE i

I REQUIREMENTS............................................ 3/4 4-26 3/4.4.d S"CCIFIC ACTIVITY (MR-UNIT 1)......................... 3f4 At 27 i

j 3/4.4.8 SPECIFICACTIVITY(FORUNIT.2tr...."....................3/4K4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC i

I ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL' POWER 1

WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >-1 pCi/ gram i

DOSE EQUIVALENT. 1-131................................... 3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS l

PR0 GRAM................................................. 3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................. 3/4 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY................................ 3/4 4-33 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 16 EFPY................................ 3/4 4-34 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM

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WITHDRAWAL SCHEDULE..................................... 3/4 4-35 Pressurizer............................................. 3/4 4-36 I

Overpressure Protection Systems......................... 3/4 4-37 i

3/4.4.10 STRUCTURAL INTEGRITY................................ 3/4 4-39 3/4.4.11 REACTOR COOLANT SYSTEM VENTS........................ 3/4 4-40 CATAWBA - UNITS 1 A 2 VII Amendment No.\\1 (Unit 1)

Amendment No.1 ( (Unit 2)

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REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 1)

All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

2)

Tubes in those areas where experience has indicated potential problems, and 3)

A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube.

If any selected tube does not permit the pas < age of the eddy current probe for a tube I

inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

c.

For Unit 1, in addition to the 3% sample, all tubes for which the alternate plugging criteria has been previously applied shall be

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inspected in the tubesheet region.

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

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partial tube inspection provided:

i 1)

The tubes selected for these samples include the tubes from l

.those areas of the tube sheet array where tubes with imperfections were previously found, and 2)

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

e.

For Unit 1, implementation of the interim steam generator tube / tube support plate elevation plugging limit fr Cy& 9 requires a 100%

l bobbin probe inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold

$5N SN"M leg tube support plate with outer diameter stress corrosion cracking ppoM ATTRRO

-(00 SCC) indications? An inspection using the rotating pancake coil PMfE:

(RPC) probe is required in order to show operability of tubes with flaw like bobbin coil signal amplitudes greater than 1.6 volt but l

less than 2.7 volts.

For tubes that will be administratively plugged l

l or repaired, no RPC inspection is required. The RPC results are to l

be evaluated to establish that the principal indications can be characterized as 00 SCC.

l The results of each sample inspection shall be classified into one of the

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following three categories:

Cateaory Insnection Results l

l C-1 Less than 5% of the total tubes inspected are degraded tabes and none of the inspected tubes are l

defective.

I CATAWBA - UNITS 1 & 2 3/4 4-13 Amendment No.

(Unit 1)

Amendment No.

(Unit 2) l l

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m REACTOR COOLANT SYSTEM i

SURVEILLANCE REOUIREMENTS (Continued)

=e 17.S' i

is verifie

+o result in total primary to secondary leakage less than -39+gpm (includes operational and accident leak-l l

age). The basis for. determining expected leak rates from j

the projected crack distribution is provided in Attachment 4 of the Supplement to Technical Specification amendment dated i

December 14, 1993 (SG-93-12-006).

l 2.

A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than 2.7 volts l,

provided a rotating pancake coil (RPC).. inspection does not 4

detect degradation.

Indications of degradation with a flaw type bobbin coil j

signal amplitude of equal to or greater than 2.7 volts will l

be plugged or repaired.

l NSERT MRNrRhpu g

N tain tubes as identified in WCAP-13494, Rev.1, will be l

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• excluded from application of the Interim Plugging Criteria Limit 9 ^C'E as it has been detarsined that these tubes may collapse or

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deform following a postulated LOCA + SSE Event.

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b.

The steam generator shall be determined OPERA 8LE after' completing the corresponding actions (plug or repair all tubes exceeding the repair

' limit and 'all tubes containing through-wall cracks) required by Table i

4.4-2.

For Unit 1, tubes with defects below F* fall under the alternate tube plugging criteria and do not have to be plugged.

I 4.4.5.5 Recorts a.

Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes repaired in each steam generator shall' be reported to tho' Commission in a Special Report i

pursuant to Specification 6.9.t; A

b.

The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to

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Specification ~6.9.2 within 12 months following the completion of the inspection. This Special Report 'shall include:

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1)

Number and extent of tubes inspected, I

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CATAWBA - UNITS 1 & 2 3/4 4-16 a Amendmeat No.

(Unit 1)

Amendment No.1 (Unit 2) 1

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REACTOR COOLANT SYSTEM SVRVElllANCE REOUIREMENTS (Continued)

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2)

Location and percent of wall-thickness penetration for each indication of an imperfection, and i

3)

Identification of tubes repaired.

For Unit 2, results of steam generator tube inspections, which fall c.

into Category C-3, shall be reported in a Special Rep;rt to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

l d.

For Unit 1, the results of inspections for all tubet for which the alternate tube plugging criteria has been applied shall be reported to the Nuclear Regulatory Commission in accordance with 10 CFR 50.4, prior to restart of the unit following the inspection. This report shall include:

1)

Identification of applicable tubes, and m

2)

Location and size of the degradation..

ha Unit 1, the results of inspections performed under 4.4.5.2 pyptAc.E 9)WM e.

all in which the tube support plate elevations inte p ugging s

gga e, criteria een applied shall be reported to theJ.omm ssion N ^*

• D following the

' ction and prior to Cycle 8 pefation. The report 9PrfrE shall include:

1.

Listing of applicable t 2.

Location (a al-a le intersections'pe e) and extent of degrad (voltage).

i Projected Steam Line Break (SLB) Leakage.

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CATAWBA - UNITS 1 & 2 3/4 4-16 b Amendment No.

(Unit 1)

Amendment No.

(Unit 2) l

D EL E TE EDT)R2 WACTORCOOLANTSYSTEM 3/ 4.8 SPECIFIC ACTIVITY (FOR UNIT 1) i LIMITING CONDITION FOR OPERATION 3.4.8 The speci activity of the reactor coolant shall be 1 ited to:

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Less than or qual to 0.58 microcurie per gram 00 EQUIVALENT I-131, j

and b.

Less than or equal o 100/E microcuries pe gram of gross specific activity.

APPLICABILITY: MODES 1, 2, 3, 4, nd 5. (Un 1)

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ACTION:

MODES 1, 2 'anB 3*:

a.

With the specific act ty of the re tor coolant greater than 0.58 l

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microcurie per gram SE EQUIVALENT I-1 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> i

during one contin us time interval or ceeding the limit line shown on Figure 3.4-be in at least HOT STAND with T,y less than 500*F within 6 hou

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b.

With the ross specific activity of the reactor oolant greater than l

100/E croCuries per gram of gross radioactivity, be in at least HOT STA Y with T,,less than 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; an i

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he provisions of Specification 3.0.4 are not applicabl

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• With T,y greater than or equal to 500*F.

CATAWBA - UNIT 1 3/4 A4-27 Amendment No. 111 (Unit 1) l

REACTOR COOLANT SYSTEM I

3/4.4.8 SPECIFIC ACTIVITY-tIOR UNIT 2)

LIMITING CONDITION FOR OPERATION

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I The specific activity of the reactor coolant shall be limited to:

3.4.8 Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, a.

and b.

Less than or equal to 100/E microcuries per gram of gross specific activity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. -(Ur.it 2) l i

ACTION:

J MODES 1, 2 and 3*:

With the specific activity.of the reactor coolant greater than 1 i

a.

microcurie per gram DOSE EQUIVALENT I-131 for mor's than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown j

on Figure 3.4-1, be in at least HOT STAN08Y with T,, less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

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b.

With the gross specific activity of the reactor coolant greater than-100/E microcuries per gram of gross radioactivity, be in at least HOT STANDBY with T,less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and l

The provisions of Specification 3.0.4 ' re not applicable.

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• With T,, greater than or equal to 500*F.

CATAWBA - UNIT 162 3/4 4-27 Amendment No. 1 (Uc.it 2) l i

REACTOR COOLANT SYSTEM BASES 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 i

for Reactor Coolant System leakage or failure due to stress corrosion.

1 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

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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 tima for taking' corrective actions to restore the contaminant concentrations to within the Steady-State Limits.

The Surveillance Requirements provide adequtte assurance that concentrations in excess of the limits will be detected in sufficient time to.*

take corrective action.

3/4.4.8 SPECIFIC ACTIVITY

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The limitations on the specific activi.ty of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 dose guideline values following a i

steam generator tube rupture accident in conjunction with an assumed steady-state primary-to-secondary steam generator leakage rate of 0.4 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are i

conservative in that specific site parameters of the Catawba site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 0.50 micr Celefgm DOSE EQUIVALENT I 131 for Uni 4-h-ed-1.0 microcurie / gram DOSE EQUIVALENT I-131 -foMMt-b but within the allow &le limit shown on Figure 3.4-1, accomodates possible iodine spiking phenonenon which may occur following changes in THEPRAL POWER.

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AmendmentNo.\\

(Unit 1)

CATAWBA - UNITS 1 & 2 B 3/4 4-5 Amendment No. 1 (Unit 2)

l Insertions for Technical Specification 3.4.5 Statement 1 l

l The determination of tube support plate intersections having OD SCC indications shall be l

based on the performance of at least 20 percent random sampling of tubes inspected over their fulllength.

4. If as a result ofleakage due to a mechanism other than OD SCC at the tube support plate intersection, or some other cause, an unscheduled mid-cycle inspection is performed, the following repair criteria apply instead of 4.4.5.4.a.13.2. If bobbin voltage is within expected limits the indication can remain in service. The expected repair limits are determined from the following equation:

At y < CL(V,-V,a)+1%

1+(.2)( At )

CL where:

V = measured voltage Vuoc = voltage at beginning of cycle (BOC)

At = time period of operation to unscheduled outage CL = cycle length (full operating cycle length where operating cycle is the time between two scheduled steam generator inspections)

Vst = 4.5 volts for 3/4 - inch tubes For implementation of the voltage-based repair criteria to tube support plate e.

intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:

1.

If the estimated leakage based on the actual measured end-of cycle voltage

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distribution would have exceeded the leak limit (for the postulated main steam line break utilizing licensing basis assumptions) during the previous operating cycle.

2.

If circumferential crack like indications are detected at the tube support plate intersections.

3.

If the indications are identified that extend beyond the confines of the tube support plate.

Page 1 of 2

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If the calculated conditional burst probability exceeds 1 X 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence.

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Page 2 of 2

m ATTACilMENT 2 i

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r TECHNICAL JUSTIFICATION Proposed Renewal Of The Interim Pluenine Criteria GPC) With'n Tech Spec 3.4.5 - Steam Generston and Increase the Soccific Activity of Tech Socc 3.4.8 l

Catawba is requesting renewal of the voltage based Interim Plugging Criteda (IPC) for Unit 1. Several items within the IPC sections of Tech Spec 3.4.5, are being revised to reflect the guidance provided in draft Generic Letter 94-XX " Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" (issued for public J

comment in the Federal Register on August 12,1994), and developments associated with other utilities in seeking approval of an interim plugging criteria. These changes include the following:.

In Tech Spec 3.4.5:

i 1)

In Surveillance Requirement (SR) 4.4.5.2.e. (page 3/4 4-13), remove reference that this criteria j

only applies to Cycle 8.

i 2)

Add statement to SR 4.5.5.2.e, (page 3/4 4-13) requiring that the determination of tube i

support plate intersections that have OD SCC will be based on performance of at least 20 l

percent random sampling of tubes inspected over their full length.

j 3)

Decrease the total primary to secondary leakage limit given in SR 4,4.5.4.a.13)l. (page 3/4 4-

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16a) from 30.0 gpm to 17.5 gpm.

4)

Insert paragraph "4" in SR 4.4.5.4.a.13 (page 3/4 4-16a) to add an allowance to modify the 1

bobbin voltage repair limit of 4.4.5.4.a.13.2 to a calculated voltage limit'in the event of an j_

unscheduled mid-cycle inspection j

5)

Replace Paragraph in SR 4.4.5.5.e. (page 3/4 4-16b) to require notifying the NRC staff prior to i

returning the steam generators to service should any of the following conditions arise:

l If the estimated leakage based on the actual measured end-of cycle voltage s

a.

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distribution would have exceeded the leak limit (for the postulated main steam line j

break utilizing licensing basis assumptions) during the previous operating cycle.

b.

If circumferential crack like indications are detected at the tube support plate intersections.

If the indications are identified that extend beyond the confines of the tube support c.

plate.

d.

If the calculated conditional burst probability exceeds 1 X 102, notify the NRC and provide an assessment of the safety significance of the occurrence.

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TECHNICAL JUSTIFICATION In Tech Spec 3.4.8:

1)

Delete entire page 3/4 A4-27. This page was the Unit 1 Specific Activity Tech Spec. The Specific Activity Tech Spec will be identical for both Unit I and Unit 2.

On Page 3/4 B4-27 delete all references to Unit 2 and the " B" reference in the page number.

Add indication that Tech Spec applies to " Units 1 & 2.

2)

In BASES 3/4.4.8, second paragraph delete, " 0.58 microcurie / gram DOSE EQUIVALENT I-131 for Unit 1, and' and " for Unit 2" i

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l TECIINICAL JUSTIFICATION Technical Justification Of Proposed Changes This license amendment request proposes the renewal of the bobbin probe voltage-based interim plugging criteria that had been previously approved for Cycle 8. Approval of this amendment will preclude unnecessary plugging or repairing tubes by sleeving due to the occurrence of outer diameter initiated stress corrosion cracking (ODSCC) at the tube support plate elevations in the Catawba Unit I steam generators. The interim plugging criteria approved for Cycle 8 and contained in the draft Generic Letter, 94-XX " Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" can be summanzed as follows:

" Flaw indications with a bobbin coil voltage less than or equal to 1.0 volt can remain in service without further action. For flaw indications in excess of 1.0 volt but less than 2.7 volts, the tube can remain in service provided an RPC inspection of the indication does not detect ODSCC or any other degradation mode. Crack indications above 2.7 volts will be plugged or repaired by sleeving, and do not require RPC confirmation."

This amendment request utilizes information previously submitted for the amendment approved for Cycle 8 and reflects the " Requested Actions" for a licensee that chooses to implement a steam generator tube interim plugging criteria, as stated in the draft Generic Letter, 94-XX " Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" This amendment request also reflects developments associated with other utilities that have received approval of an interim plugging criteria.

The changes being proposed to the Tech Specs do not alter the interim plugging criteria currently stated in Tech Specs which was approved and utilized during Cycle 8. The primary change to the Tech Specs is to incorporate the guidance of draft Generic Letter, 94-XX " Voltage-Based Repair Criteria

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for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" which will allow removal of the cycle specific limitation currently in the Tech Spec.

Each of the " Requested Actions" of the draft Generic Letter is addressed below:

1)

Implementation ofthe inspection guidance discussedin Section 3 ofEnclosure 1. 7he inspection guidance ensures that the techniques used to inspect steam generators are consistent with the techniques used to develop voltage-basedrepair limit methodology.

Page 3 of 9

TECIINICAL JUSTIFICATION Catawba is committing to incorporate the inspection guidance provided in this section of the draft Generic Letter into the steam generator inspection program, with the exception of bobbin coil calibration and probe wear re-inspection requirements.

The Generic Letter requires the eddy current normalization to be performed using four 100% through wall holes instead of Catawba's current method of four 20% through wall flat bottom holes.

Catawba has used the 20% holes since IPC was implemented in 1991. Transfer standards have also been used to adjust our voltages to the standards used in the industry database. Changing the normalization method would introduce more variability in the Catawba data. Catawba's current inspection guidelines are consistent with the inspection guidelines of Appendix A of Reference 2, which is a stated requirement of the Tech Spec.

Therefore, Catawba will continue the current normahzation method of four 20% through wall flat bottom holes.

The Generic Letter also requires that when utilizing the periodic wear measurement approach, if a probe is found out of specification, all tubes inspected since the last successful calibration should be re-inspected with the new calibrated probe. Compliance with this standard is impracticable and unnecessarv. It is impracticable because wear standard data acquisition has produced erratic results in the field, causing new probes to be rejected along with those probes used for arbitrary numbers of tubes. Data was presented in the NEI/NRC meeting on Novcmber 3,1994, which concluded that for flaws greater than 0.5 volts, there is no loss of detection prcbability for all degrees of wear, down to complete loss of centering.. Smaller signals (0.2 volts) ccrresponding to a 20% EDM notch were detectable but not always distinguishable from the noise. For a 1 volt indication, the voltage ranged from.7 to 1.5 volts. Since detection has been achieved without reference to wear behavior, Catawba will retest with a probe controlled to the wear standard variability limit, all OD SCC indications exceeding 0.7 volts. The purpose of this retest 6 Westablit the correct signal amplitude.

Although no specific requirements exist regarding nominal probe size, Catawba intends to utilize a 0.630 inch diameter bobbin probe in lieu of a 0.610 inch diameter bobbin probe. This is a change from Catawba Unit 1, Cycle 8 practice.

2)

Calculation ofthe leakageper the guidance ofSection 2.b ofEnclosure 1. 7his calculation, in conjunction with the use oflicensing basis assumptionsfor calculating offsite releases, enables the licensees to demonstrate that the applicable limits of 10 CFR 100 continue to be met.

l The Generic letter further states that a methodology should be submitted for the NRC review and l

approval for calculating the total primary-to-secondaiy leak rate in the faulted steam generator during a postulated MSLB assumed to occur at EOL. The methodology for calculating total leak rate that will be utilized at Catawba is the Full Monte Carlo method. This method is described in Reference 9, l

Section 6.8.1.

This methodology would be acceptable under the draft Generic Letter and was approved by the NRC in Reference 11.

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TECIINICAL JUSTIFICATION 3)

Calculation ofconditional burstprobabilityper the guidance ofSection 2.a of. 7his is a calculation to assess the voltage distribution left in service against a threshold value.

The methodology for calculation of conditional burst probability is also required to be reviewed and approved by the NRC as part of the Generic Letter requirements for implementing IPC. Catawba intends to use a methodology similar to the methodology employed for determining leakage. This methodology, similar to the leakage methodology, will account for uncertainties in tk ccslation parameters. A description of this methodology will be submitted to the NRC for review and approval at a later date, expected to be in December 1994.

4)

Implementation ofthe operatiornileakage limits identifiedin Section 5 of. The operationallimit is a defense-in-depth measure thatprovides e meansfor identifying leaks during operation to enable repir before such leaks result in tubefailure.

Catawba is currently in compliance with this requirement of the Generic Letter. Implementation of the operational leakage limit of 150 gpd through each steam generator was implemented by Tech Spec Amendment (No.102/96) for Cycle 7.

5)

Review ofthe leakage monitoring measures including theproceduresfor timely detection, trending, andresjxinse to rapidlyincreasingleaks. The objectim is to ensure that shoulda sigmficant leak be experiencedin service, it will be detectedand theplant shut down in a timely manner to reduce the likelihoodofapotentialrupture.

Catawba is in compliance with this requirement of the Generic Letter. Catawba has previously completed review of the leakage monitoring measures and procedures for timely detection, trending and response to rapidly increasing leaks. The results of this review was submitted to the NRC (in Reference 1 and reviewed by the NRC as part of Catawba's request for IPC Tech Spec Amendment (No. I11/105) for Cycle 8.

6)

Accptisition of the tubepulldataper the guidance ofSection 1 ofEnclosure 1. It is necessary to acquirepulled tube data to cortfirm the degradation mechanism.

Catawba has met the intent of this requirement from past tube pulls. Beginning in 1990, nine tubes have been pulled from Catawba Unit 1, representing 18 intersections. These tube pulls have confumed the degradation mechanism and provided additional data to the databases. This proactive approach has further educated both Duke Power and the industry. Catawba recognizes that our steam generators are degraded and have significantly impacted our nuclear station performance. The alternate repair Page 5 of 9

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TECllNICAL JUSTIFICATION criterion is part of our short term strategy while our long term strategy is to replace the steam generators. The replacement is currently scheduled for the EOC-9 in 1996 at Catawba Unit 1.

1 Additional tube pulling is not necessary to implement IPC for Catawba Unit 1.

l 7)

Reporting restdtsper the guidance ofsection 6 ofEnciasure 1.

Catawba commits to reporting results as required by the Generic Letter. This Tech Spec Amendment Request reflects the reporting requirements of the Model Tech Spec as presented in Enclosure 2 of the draft Generic Letter.

i 8)

Submittal ofa technical specyication (IS) amendment request that commits to the i

above aidprovides ISpagesper the guidance ofEnclosure 2 including the associated consideration ofno sigmficant ha:ards consideration (10 CFR 50.92) and l

supportingsafety analysis.

1 Most of the requirements of the draft Generic Letter have been previously submitted to the NRC as a Tech Spec Amendment Request (References 1 - 7) for Cycle 8 which was approved by the NRC.

(Reference 8). Additional requirements of the Generic Letter beyond what was approved for Cycle 8 i

and contained in Enclosure 2 of the draft Generic letter (Model Tech Spec), are submitted with this l

Tech Spec Amendment Request.

The most significant numerical changes in this Tech Spec Amendment Request are a decrease in l

allowable SLB leakage from 30.0 gpm to 17.5 gpm. This decrease in allowable SLB leakage is in support ofincreasing the Tech Spec limit on Reactor Coolant Specific Activity back to 1.0 microcurie / gram DOSE EQUIVALENT I-131. The Tech Spec Amendment for Cycle 8 that was approved, required that several data points that were excluded from the database to determine the repair voltage limit, be included in the database (herein referred as the NRC database) for Cycle 8 IPC application. Also, the primary-to-secondary leakage was calculated in accordance with the methodology given in draft NUREG-1477, which does not recognize a voltage-to leakage correlation.

As a result of the inclusion of these data points, utilizing the NUREG 1477 methodology, and utilizing the EOC-7 steam generator tube inspection results to project the EOC-8 conditions, the leakage w~as calculated to be 29.35 gpm. The dose calculation (Reference 7) submitted as part ofReference 5, bounded a leakage rate of 17.5 gpm with a Reactor Coolant Specific Activity of 1.0 microCuries/ gram DOSE EQUIVALENT I-131. In order to maintain dose consequences of a SLB a small fraction of 10 CFR 100 limits, the Reactor Coolant Specific Activity Tech Spec was reduced to 0.58 microCuries/ gram DOSE EQUIVALENT I-131 for Cycle 8.

I Since the approval of the Cycle 8 IPC Tech Spec Amendment, it has been adequately demonstrated that there is a correlation between leak rate and bobbin voltage, and the NRC has accepted this correlation in the Braidwood-l IPC SER (Reference 10). The NRC has also recognized (in the Braidwood 1 SER, Reference 11) that there is sufficient bases to exclude the outlier data points that Page 6 of 9 i

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TECHNICAL JUSTIFICATION were required to be included by the Catawba 1 Cycle 8 IPC NRC SER (Reference 8). However, using the accepted leak rate versus bobbin voltage correlation methodology (full Monte Carlo) for calculating leak rate with the "NRC in bse" and the NRC required POD of 0.6, the SLB leak rate is calculated to be l.61 gpm for EOC-8 projected voltage distribution. Using the EPRI database i

(excludes the outlier data points), with a POD of 0.6, the resultant leak rate is 0.36 gpm. Both of these leak rates are well within the bounding leak rate of 17.5 gpm. Reference 12 provides detailed discussion of these leak rate calculations. It should be noted that this methodology and database is acceptable under the NRC draft Generic Letter, 1

I Based on the calculated leak rates presented above being bounded by the dose calculation (Reference

7) submitted in Reference 5, the Tech Spec for Reactor Coolant Specific Actisity can be raised back to 1.0 microCuries/ gram DOSE EQUIVALENT I-131. with the dose consequences for a SLB remaining below a small fraction of 10 CFR 100 limits.

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I TECIINICAL JUSTIFICATION References l.

Technical Specification Amendment Request for Catawba Unit 1, Letter to NRC dated October 5,1993, Docket No. 50-413, " Renewal of Steam Generator Tube Interim Plugging Criteria for Unit 1 Cycle 8" 2.

WCAP-13854, " Technical Support for Cycle 8 Steam Generator Tube Interim Plugging Criteria For Catawba Unit 1," September 1993.

3.

Supplement to Technical Specification Amendment Request for Catawba Unit 1, IAter to NRC dated October 14, 1993, Docket No. 50-413, " Renewal of Steam Generator Tube Interim Plugging Criteria for Unit 1 Cycle 8" 4.

Supplement to Technical Specification Amendment Request for Catawba Unit 1, Letter to NRC dated October 28, 1993, Docket No. 50-413, " Renewal of Steam Generator Tube Interim Plugging Criteria for Unit 1 Cycle 8" 5.

Supplement to Technical Specification Amendment Request for Catawba Unit 1, Letter to NRC dated December 14, 1993, Docket No. 50-413 and 50-414, " Renewal of Steam Generator Tube Interim Plugging Criteria for Unit 1 Cycle 8" 6.

Supplement to Technical Specification Amendment Request for Catawba Unit 1, Letter to NRC dated December 20,1993, Docket No. 50-413, " Renewal of Steam Generator Tube Interim Plugging Criteria for Unit 1 Cycle 8" 7.

Duke Power Calculation No. CNC-1227.00-00-0051, Rev. 3, "Offsite Dose From A Postulated Main Steam Line Break.

8.

Safety Evaluation by the Office ofNuclear Reactor Regulation Related to Amendment No. I11 to Facility Operating License NPF-35 and Amendment No.105 to Facility Operating License NPF-52, Duke Power Company, ET AL., Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 50-414, dated December 16,1993.

9.

WCAP-14046, Revision 1, "Braidwood Unit 1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria" 10.

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 50 to Facility Operating License NPF-72, Commonwealth Edison Company, Braidwood Station, Unit 1, Docket No. STN 50-456, dated May 7,1994.

Page 8 of 9

TECHNICAL JUSTIFICATION 11.

Safety Evaluation by the Office ofNuclear Reactor Regulation Related to Amendment No. 54 to Facility Operating License NPF-72, Commonwealth Edison Company, Braidwood Station,

.I Unit 1, Docket No. STN 50-456, dated August 18,1994.

l NOTE:

All of the above references have been previously submitted to the NRC.

The following Reference is being submitted with this request.

12.

Catawba - 1, Cycle 7 IPC Assessment and Projected EOC-8 SLB leakage.

dated November,1994 Page 9 of 9

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NO SIGNIFICANT IIAZARDS CONSIDERATION i

In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license j

amendment is analyzed using the following standards and found not to: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a i

new or d;fferent kind of accident from any accident previously evaluated; or 3) involve a significant reduction in margin ofsafety.

Conformance of the proposed amendment to the standards for a determination of no significant hazard

)

as defined in 10 CFR 50.92 (three factor test) is shown in the following:

1)

Operation of Catawba Unit 1 in accordance with the proposed licen.ce amendment does not involve a significant increase in the probability or consequences of an accident l

previously evaluated.

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A single tube rupture is not anticipated during operation of Catawba Unit 1. Based on the existing data j

base, the limiting RG 1.121 criterion for tube burst capability of 3 times normal operating differential is l

satisfied with 3/4" diameter tubing with bobbin coil indi:ations with signal amplitudes less than 4.54 volts, regardless of the indicated depth measurement. This structural limit is based on a lower 95%

l prediction bound of the data and using LTL material propenies A 1.0 volt plugging criteria compares favorably with the structural limit considering the previously calculated growth rates for ODSCC l

within the Catawba Unit I steam generators. Assuming a voltage increase of 0.4 volts, and adding a l

l 14% NDE uncenainty of 0.14 volts (90% cumulative probability) to the interim plugging criteria of 1.0 l

volt results in an EOC voltage of approximately 1.6 volts. This end of cycle voltage compares favorably with the Stmetural Limit of 4.54 volts. The applicability of assumed growth rates for each cycle of operation will be confirmed prior to retum to power of Catawba Unit 1. A similar stmetural margin is anticipated for subsequent cycles.

In addition, for an EOC voltage structural limit of 4.54 volts, applying the 40% growth allowance and the 14% NDE uncertainty results in a margin between the structural limit and the attemate repair limit (2.7 volts), which is well within the structural limit. This repair limit will be applied for IPC implementation to repair bobbin indications greater than 2.7 volts independent of RPC confirmation of the indication.

Concerning SLB leakage in support of implementation of the interim plugging criteria, it will be determined whether the distribution of cracking indications at the tube support plate intersections at the end of a cycle are projected to be such that primary to secondary leakage would result in site boundary doses within the pertinent 10 CFR 100 limits.

The SLB leakage rate calculation methodology l

prescribed in Reference 2 will be used to calculate End of Cycle SLB leakage. Based on EOC 8 l

projections, it is calculated (Reference 3) that leakage during a postulated SLB event at the EOC 8 will l

be limited to approximately 1.61 gpm which is shown to result in acceptable dose consequences (Reference 1). Reference 1 shows that SLB leakage of 17.5 gpm in the faulted loop results in dose i

consequences which are less than the pertinent 10 CFR 100 limits. Similar results are expected for Page 1 of 6 l

1 NO SIGNIFICANT HAZARDS CONSIDERATION subsequent cycles and confirmation of leak rates will be performed prior to placing the Steam generators in service.

Therefore, renewal of the proposed 1.0 volt interim plugging criteria does not adversely affect steam generator tube integrity and results in acceptable dose consequences. The proposed amendment does not result in any increase in the probability or consequences of an accident previously evaluated within the Catawba Unit 1 FSAR.

2)

The proposed license amendment does not citate the possibility of a new or different kind of accident from any accident previously evaluated.

Renewal of the proposed steam generator tube interim plugging criteria does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism which could result in an accident outside of the region of the tube support plate elevations - no ODSCC is occurring outside the thicL:ess of the tube support plates. Neither a single or multiple tube rupture event would be expected in a steam generator in which the plugging criteria has been applied (during all plant conditions).

Upon application of the interim plugging criteria, no primary to secondary leakage during normal operation is anticipated during all plant conditions due to degradation at the tube support plate elevations in the Catawba Unit I steam generators. However, additional conservatism is built into the existing operating leakage limit with regard to protection against the maximum permissible single crack l

length which may be achieved during operation due, in large part, to the potential occurrence of through-wall cracks at locations other than the tube support plate intersections.

Application of the 1.0 volt interim steam generator tube plugging criteria at Catawba Unit 1 is not expected to result in tube burst during all plant conditions during operation. Tube burst margins are expected to meet RG 1.121 acceptance criteria. The limiting consequence of the application of the interim plugging criteria is a potential for SLB leakage. The methodology for calculating SLB leak rate uses a voltage-to-leakage correlation and this methodology has previously been resiewed and approved by the NRC. The SLB leakage value will be confirmed to be less than allowable levels prior to return l

to power of Catawba Unit 1. No unacceptable leakage is anticipated at normal operating or RCP locked rotor conditions.

Therefore, as the existing tube integrity criteria and accident analyses assumptions and results will continue to be met, the proposed license amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

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l NO SIGNIFICANT IIAZARDS CONSIDERATION l

i 3)

The proposed license amendment does not involve a significant reduction in margin of I

safety.

The use of the voltage based bobbin probe interim tube support plate eJevation plugging criteria at l

Catawba Unit 1 is demonstrated to maintain steam generator tube integrity commensurate with the i

criteria of Regulatory Guide 1.121. RG 1.121 describes a method acceptable to the NRC staff for meeting GDCs 14,15, 31, and 31 by reducing the probability or the consequences of steam generator tube rupture. This is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established by insenice inspection, for which tubes with unacceptable cracking should be removed from senice. Implementation of the bobbin probe voltage based interim tube l

plugging criteria of 1.0 volt is supplemented by enhanced eddy current inspection guidelines to provide consistency in voltage normalization, a 100% eddy current inspection at the tube support plate elevations, and rotating pancake coil inspection requirements for the larger indications left in service to l

characterize the principle degradation as ODSCC. Even under the worst case conditions, the occurrence ofODSCC at the tube support plate elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions.

Based on the analyses for Cycle 8, the expected leakage values and the leakage conditions required to be confirmed during accidents creating high differential pressures across the steam generator tubes (e.g., SLB), dose analysis confinn the maximum permissible leakage will result in offsite dose consequences within the guideline values. A MSLB accident with assumed leakage growth in the faulted generator results in the EAB and LPZ doses remaining within 10% of the 10 CFR 100 values of 25 Rem whole body and 300 Rem thyroid for the accident-initiated iodine spike, and 10 CFR 100 values for the pre-accident iodine spike (Reference 1).

The distribution of crack indications at the tube support plate elevations will be confirmed to result in acceptable primary to secondary leakage during all plant conditions and that radiological consequences are not adversely impacted.

Renewal of the tube support plate elevation plugging criteria for operation at Catawba Unit I will decrease the number of tubes which must be repaired by sleeving or taken out of service by plugging.

The installation of steam generator tube plugs reduce the RCS flow margin. Thus, implementation of the altemate plugging criteria will maintain the margin of flow that would otherwise be reduced in the event ofincreased tube plugging.

Based on the above, it is concluded that the proposed license amendment request does not result in a i

significant reduction in margin with respect to plant safety as defined in the Final Safety Analysis l

Report or any Bases of the plant Technical Specifications.

Page 3 of 6

NO SIGNIFICANT HAZARDS CONSIDERATION CONCLUSION Based on the preceding analysis, it is concluded that using voltage-based interim steam generator tube plugging criteria for removing tubes from service and the increase in the allowable iodine activity in Tech Spec 3.4.8 for Catawba Unit 1 is acceptable and the proposed license amendment involves a no significant hazards consideration as defined in 10 CFR 50.92.

Page 4 of 6

NO SIGNIFICANT IIAZARDS CONSIDERATION References 1.

Duke Power Calculation No. CNC-1227.00-00-0051, Rev. 3, "Offsite Dose From A Postulated Main Steam Line Break.

2.

WCAP-14046, Revision 1, "Braidwood Unit 1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria" 3.

Catawba - 1, Cycle 7 IPC Assessment and Projected EOC-8 SLB leakage.

dated November,1994 Page 5 of 6

ENVIRONMENTAL IMPACT STATEMENT -

The proposed amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. As described above, the proposed amendment does not involve any significant hazards consideration, nor a significant increase or change in the types or amounts of efiluents that may be released offsite, nor a significant increase in the individual or cumulative occupational radiation exposures. Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement.

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