ML20092E283

Proposed Tech Specs Re Alternate Plugging Criteria for SG Tubing

ML20092E283
Person / Time
Site:	Sequoyah
Issue date:	09/07/1995
From:	TENNESSEE VALLEY AUTHORITY
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ML20092E278	List: ML20092E274 ML20092E283
References
NUDOCS 9509150140
Download: ML20092E283 (46)

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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNIT 1 DOCKET NO. 50-327 (TVA-SON-TS 95-15, REVISION 1)

LIST OF AFFECTED PAGES Unit 1 3/4 4-7 3/4 4-9 3/4 4-10 3/4 4-14 8 3/4 4-3 8 3/4 4-4 l

9509150140 950907 l PDR ADOCK 05000327 P - . .

. - . . _ .. . _. P D R _ _ ,

i

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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

[ The tubes selected as the second and third samples (if required by c.

Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfec-tions were previously found.

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

NOTE: Tube degradation identified in the portion of the tube that R19,.

is not a reactor coolant pressure boundary (tube end up to T" '"

• the start of the tube-to-tubesheet weld) is excluded from the Result and Action Required in Table 4.4-2.

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

Cateoory Insoection Results 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. l C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

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

' Anendnent No. 189 SEQUOYAH - UNIT 1 3/4 4-7 October 20, 1994

l t

RFACTOR COOLANT SYSTEM

+

{ SURVEit.tANCE REOUTREMENTS (Continued) l i i l

4.4.5.4 Ac'centance Criteria l a. As used in this Specification:

I Imoerfection means an exception to the dimensions, finish or

1.

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

2.

Dearadation means a service-induced cracking,

wastage, wear or i

general corrosion occuring or either inside or outside of a i tube.

J 3. Dearaded Tube means a tube containing imperfections greater l

than or equal to 20% of the nominal wall thickness caused by ,

degradation.

4.  % Deoradation means the percentage of the tube wall thickness

affected or removed by degradation.

1

5. Defect means an imperfection of such severity that it exceeds l

j the plugging limit. A tube containing a defect is defective. I Pluaaina limit means the imperfection depth at or beyond.

i 6.

which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness. Plugging limit does

not apply to that portion of the tube that is not within the pressure boundary of the reactor coolant system (tube end up to the start of the tube-to-tubesheet weld) y Lf M

7. Unserviceable describes the condition of a tube if it leaks or contains _ a defect large enough to affect its structural j 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. Tube Insoection means an inspection of the steam generator tube i

from the point of entry (hot leg side) completely around the U-bend to the top support of the cold lag.

9. Preservice insoection means a tube inspection of the full length of each tube in each steam generator performed by eddy

' current techniques prior to service establish a baseline con-

' dition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

j Q+

Amendment No. 109 3/4 4-9 october 20, 1994 SEQUOYAH'- UNIT 1 ,

i REACTOR COOLANT SYSTEM

-SURVEILLANCE REQUIREMENTS (Continued)

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged 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 submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following completion of the inspection. This Special Report shall include:

1. Number and extent of tubes inspected.

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

3. Identification of tubes plugged.

. Results of steam generator tube inspections which fall into Category E C-3 shall be reported pursuant to Specification 6.6.1 prior to resump-tion 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.

Tn 2v W E ,

4 tes ,

November 23, 1984 3/4 4-10 Amendment No. 36 SEQUOYAH - UNIT 1

F

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

a. No PRESSURE BOUNDARY LEAKAGE,

.rM b. 1 GPM UNIDENTIFIED LEAKAGE,

-f g c. CP" t;t ? pr'::ry-t; ::::nderj le:h:g: thrcq  ?' ::::: g:n: ret:r; ;,d'"'

g:!?:n: per d:y thr::;h :ny :n: ;;;;; ;;n:r:: r,

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

% 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 + 20 psig. I

f. 1 GPM leakage at a Reactor Coolant System pressure of 2235 2 20 psig ln t from any Reactor Coolant System Pressure Isolation Valve specified in j Table 3.4-1.

APPLICABILITY: H0 DES 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 I

above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage f rom

( -

Reactor Coolant System Pressure Isolation Valves, 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 STANOBY 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 f ollowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l

c. With any Reactor Coolant System Pressure Isolation Valve leakage I greater than the above limit, isolate the high pressure portion of '

the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by

. ' us'e of at least two closed manual or deactivated automatic valves.

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, hours.

,-  :::;. SURVEILLANCE REQUIREMENTS l

4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within u each of the above limits by:

f.! -

MAR 261982 SEQUOYAH - UNIT I 3/4 4- 14 Amendment No.Id

REACTOR COOLANT SYSTEM 4

BASES The plant is expected to be operated in a manner such that the secondary 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 limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during 150 plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage - 4004 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit durin operation will have an adequate margin of safety to withstand th oads imposed during normal operation and by postulated accidents./"Oper:t;ng phnts hwc demonstrated that primary-to-secondary leakage of-500 gallons per day per steam generator can readily be detected.by radiation monitors of steam s3,0d' 6s generator blowdowng Leakage in excess of this limit will require plant during w_hich the leaking tubes will be shutdown incated and and an unscheduled plugged. M inspection,Qff. is

)

G Wastage-type defects are unlikely with proper chemistry treatment of the

1. ve. secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

< e lr li,Q Plugging will be required for all tubes with imperfections exceeding theof the The portion L['p"'g* plu;;ing l!=it cf 40% ef the tcbc nc=in:1 =11 thickncss.

Smdhnce. tube is not thatthe within theRCS plugging limit pressure does not boundary apply (tube to to end up is the thestart portion oftube-of the the tube RIO tha a The tube end to tube-to-tubesheet weld portion of the tube g Reg ntM t to-tubesheet weld).g9,gy, does not affect structural integrity indications found in this portion of the tube will be excluded from the Result and Action Required for tube inspections. It is expected that any indications that extend from this region will be detected during the scheduled tube inspections. 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.

. Insert Whenever the results of any steam generator tubing inservice inspection g

m e ,_ fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.6.1 prior to resumption of plant opera- R4 <

tion. 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.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS ,

i

The RCS leakage detection systems required by this specification are i

provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of

' Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection i Systems," May 1973. l 1

l

' Amendment No. 36, 189 SEQUOYAH - UNIT 1 B 3/4 4-3 October 20, 1994

_ _ _ _ _ _ . ~ .___ _ _ _ . _ _ _ _ _ _ . _ _ . _ _ _ _ _ . ._ _

b REACTOR COOLANT SYSTEM l BASES  ;

3/4.4.6.2 QPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced ,

to a threshold value of less than 1 GPM. This threshold value is sufficiently, low to ensure early detection of additional leakage.

The surveillance requirements for RCS Pressure Isola *.fon Valves provide added assurances of valve integrity. thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS isolation valves is ' IDENTIFIED LEAKAGE and will be considered as a portion of the allowed FP limit.

i The.10. GPM IDENTIFIED LEAKAGE Ifmitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with j

the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.

This limitation ensures that in the event of a LOCA, the safety injection flow Ia'[ 11 not be less than assumed.in the accident analyses.

6/ The tot steam generator be leakage limit f 1 GPM for all s am j generators ens es that the dosag contribution fro the tube leakage 'll be i limited to a sma fraction of Part 100 limits in the vent of either a eam  ;

enerator tube rup re or steam line eak. The 4-GPM-

• it is consistent

w h the assumptions sed in the analys s of these accide s. The 400 gpd ]

lea e limit per ste generator ensure that steam genera r tube integrity is ma tained in the eve of a main steam ine rupture or un r LOCA conditions.

l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may

be indicative of an impending gross failure of the pressure boundary. Therefore, l

, the presence of any PRESSURE BOUNDARY LEAXAGE requires the unit to be promptly placed in COLD SHUTOOWN.

3/4.4.7 CHEMISTRY

^

The limitations on Reactor. Coolant System chemistry ensure that corrosion of the Reactor Coolant System'is minimized and reduces the potential for

Reactor Coolant System leakage or failure due to stress corrosion. Maintaining i- the chemistry within the Steady State Limits provides adequate corrosion

, protecti_on 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 i

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 concentra-tions to within the Steady State Limits.

SEQUOYAH - UNIT 1 'B 3/4 4-4 SEP 171980 September 17, 1980 J

---

A-A---------- .v --.-n -,<v,.,---n- - - - . - , , , , e-, , --

n-e-<

insert A i

4.4.5.2.b.4 Indications left in service as a result of application of the tube support

' plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

a l

i 5 Insert B 4.4.5.2.d. Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg 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.

Insert C 4.4.5.4.a.6. This definition does not apply to tube support plate intersections if the voltage-based repair criteria are being applied. Refer to 4.4.5.4.a.10 for the repair limit applicable to these intersections.

l 4

Insert D 4.4.5.4.a.10 Tube Sucoort Plate Pluaaina Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support  !

plate intersections, the plugging (repair) limit is based cn maintaining l I

steam generator tube serviceability as described below-I

a. Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube 4 support plate with bobbin voltages less than or equal to the lower voltage repair limit (Note 1), will be 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 the lower voltage repair limit (Note 1), will be repaired or plugged, except as noted in 4.4.5.4.a.10.c below.

7

4 insert D (continued) 4

c. Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the r bounds of the tube support plate with a bobbin voltage greater than the lower volta 0e repair limit (Note 1), but less than or equal to upper l voltage repair limit (Note 2), may remain in service if a rotating i pancake coil inspection does not detect degradation. Steam generator i

tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin coil voltage greater than the upper voltage repair limit (Note 2) will be plugged or repaired.

d. Not applicable to SON.

e. If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in

< 4.4.5.4.a.10.a,4.4.5.4.a.10.b, and 4.4.5.4.a.10.c. ,

The mid-cycle repair limits are determined from the following equations:

Va V"" = 1.0 + NDE + Gr ( )

Vuta = Vuua-(Va-Vt s) ( )

L where:

l

. Vua = upper voltage repair limit

, Vos = lower voltage repair limit l Vuum = mid-cycle upper voltage repair limit based on time into cycle I

Vuta = mid-cycle lower voltage repair limit based on Vuus and time into cycle At = length of time since last scheduled inspection dunng which Vua and Vt g were implemented cycle length (the time between two scheduled CL =

steam generator inspections)

, Vst = structural limit voltage Gr = average growth rate per cycle length NDE = 95-percent cumulative probability allowance

for nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC) i' Implementation of these mid-cycle repair limits should follow the same approach as in TS t 4.4.5.4.a.10.a,4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.

. Note 1: The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing or 2.0 volts for 7/8-inch diameter tubing.

Note 2: - The upper voltage repair limit is calculated according to the methodology in GL 95 05 as supplemented. Va may differ at the TSPs and flow distribution baffle.

Insert E 4.4.5.5.d For implementation of the voltage-based repair criteria to tube support 1 plate intersections, notify the staff prior to returning the steam generators  !

to service should any of the following conditions arise:

1. If estimated leakage based on the projected end-of-cycle (or if not practical using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break ) for the next operating cycle.

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

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

4. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end of-cycle) voltage distribution exceeds 1 x 102, notify the NRC and provide an assessment of the safety significance of the occurrence, insert F 1

3.4.6.2.c 150 gallons per day of primary to-secondary leakage through any one steam generator.

! Insert G The voltage-based repair limits of SR 4.4.5 implement the guidance in GL 95-05 and are applicable only to Westinghouse-designed steam generators (S/Gs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to tube support plate intersections. The voltage-based repair limits are not applicable to other forms of S/G tube degradation nor are they applicable to ODSCC that occurs at other locations within tho S/G. Additionally, the repair criteria apply only to indications where the degradation me hanism is dominantly axial ODSCC with no significant cracks extending outside the thicl mess of the support plate. Refer to GL 95-05 for additional description of the degradation morphology.

Implernentation of SR 4.4.5 requires a derivation of the voltage structurallimit from the i . burst versus volteqe empirical correlation and then the subsequent derivation of the I voltage repair limit kom the structural limit (which is then implemented by this j surveillance).

i l

l Insert G (continued)

The voltage structurallimit 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 percent tolerance bound for tubing material properties at 650*F (i.e., the 95 percent LTL curve). The voltage structurallimit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty. The upper voltage repair limit: Vua, is determined from the structural voltage limit by applying the following equatiw:

Vua = Vst - Von - V, where Von represents the allowance for flaw growth between inspections and Va represents the allowance for potential sources of error in the measurement of the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.

The mid-cycle equation of SR 4.4.5.4.a.10.e should only be used during unplanned inspection 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 95-05 for situations which NRC wants to be notified prior to returning the S/Gs to service. For SR 4.4.5.5.d, items 3 and 4, indications are applicable only where alternate plugging criteria is being applL.d. For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the S/Gs to service. Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing GL Sections 6.a.1 and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per GL Section 6.b(c) criteria.

Insert H Tubes experiencing outside diameter stress corrosion cracking within the thickness of the tube support plate are plugged or repaired by the criteria of 4.4.5.4.a.10.

Insert i The total steam generator tube leakage limit of 600 gallons per day for all steam l generators and 150 gallons per day for any one steam generator will minimize the  !

potential for a significant leakage event during steam line break. Based on the NDE l uncertainties, bobbin coil voltage distribution and crack growth rate from the previous I inspection, the exnected leak rate following a steam line rupture is limited to below )

4.3 gpm in the faulted loop, which will limit the calculated 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 4.3 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed from serWee in order to reduce the postulated primary-to secondary steam line break leakage to below 4.3 gpm.

u. a ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNIT 1 DOCKET NO. 50-327 (TVA-SON-TS-95-15, REVISION 1)

DESCRIPTION AND JUSTIFICATION FOR TS AMENDMENT' 4

l

)

4 i

/

4 i

Descriotion of Chanae TVA proposes to modify the SON Unit 1 technical specifications (TSs) to incorporate

! new requirements associated with steam generator (S/G) tube inspection and repair. The new requirements establish alternate S/G tube plugging criteria at tube support plate ,

(TSP) intersections. The proposed changes are as follows: l

~

l

1. Add Surveillance Requirement (SR) 4.4.5.2.b.4 j

" Indications left in service as a result of application of the tube support plate

. voltage based plugging repair criteria shall be inspected by bobbin coil probe during all future refueling outages."

2. Add SR 4.4.5.2.d.

P

" Implementation of the steam generator tube / tube support plate repair criteria

, requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support j plate intersections down to the lowest cold-leg tube support plate with known

! outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications i shall be based on the performance of at least a 20 percent random sampling of tubes j j inspected over their fulllength." l

3. Add requirements to SR 4.4.5.4.a.6.

"This definition does not apply to tube support plate intersections if the i voltage-based repair criteria are being app'ied. Refer to 4.4.5.4.a.10 for the repair l limit applicable to these intersections._

4. Add SR 4.4.5.4.a.10 -

" Tube Sucoort Plate Pluooino Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of 4 the tube support plates. At tube support plate intersections, the plugging (repair) l

! limit is based on maintaining stearn generator tube serviceability as described below:

a. Steam generator tubes, whose degradation is attributed to outside diameter

,F stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to the lower voltage repair limit (Note 1), will be allowed to remain in service.

l 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 the lower voltage repair limit (Note 1), will be i repaired or plugged, except as noted in 4.4.5.4.a.10.c below.

1

2-

c. Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (Note 1),

but less than or equal to upper voltage repair limit (Note 2), may remain in service if a rotating pancake coilinspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin coil voltage greater than the upper voltage repair limit (Note 2) will be plugged or repaired.

d. Not applicable to SON.

e. If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.a.10.a, b, & c.

The mid-cycle repair limits are determined from the following equations:

Vst V"" = 1.0 + NDE + Gr ( ).

Vuta = Vuua-(Vua-Vt a) ( )

where:

Vav

= upper voltage repair limit I Vat

= lower voltage repair limit V uus = mid-cycle upper voltage repair limit based on time into cycle Vuta = mid-cycle lower voltage repair limit based on Vuua and time into cycle At = length of time since last scheduled inspection during which Vua and Vt a were implemented CL = cycle length (the time between two scheduled steam generator inspections)

Vst = structural limit voltage Gr = average growth rate per cycle length NDE = 95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC)

e J

4 Implementation of these mid-cycle repair limits should follow the same approach as in 2 TSs 4.4.5.4.a.10.a,4.4.5.4.a.10.b, and 4.4.5.4.a.10.c."

\

Note 1: The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing or l 2.0 volts for 7/8-inch diameter tubing.

Note 2: The upper voltage repair limit is calculated according to the methodology in l GL 95-05 as supplemented.' Va may differ at the TSPs and flow distribution baffle.

5. Add SR 4.4.5.5.d.

"For implementation of the voltage-based repair criteria to tube support plate intersections, notify the staff prior to returning the steam generators to service should any of the following conditions arise:

i

1. If estimated leakage based on the projected end-of-cycle (or if not practical using 1 the actual measured end-of-cycle) voltage distribution exceeds the leak limit l

(determined from the licensing basis dose calculation for the postulated main steam line break ) for the next operating cycle.

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

2

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

l

4. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution i exceeds 1 x 10'2, notify the NRC and provide an assessment of the safety significance of the occurrence."

6. Replace SR 3.4.6.2.c with

" Primary-to-secondary leakage shall be limited to 150 gallons per day through any one steam generator "

7. Change Bases 3/4.4.5, " Steam Generator," to reflect the new primary-to-secondary leakage limit (150 gallons per day per S/G) and include a reference to the tube repair limit as defined in Specification 4.4.5.4.a. In addition, Bases Section 3/4.4.6.2,

" Operational Leakage," is revised to reflect the S/G operationalleakage limits. l

Reason for Chanae TVA is proposing to change SON Unit 1 TSs to reduce the need for repairing or plugging S/G tubes having indications that exceed the current TS depth-based plugging limit.

TVA proposes to add alternate tube plugging criteria at TSP intersections that are based on maintaining structural and leakage integrity of tubes with indications of ODSCC within the confines of the TSP regions. Westinghouse Electric Corporation has performed analyses to: (1) show that indications within the TSP region meet Regulatory Guide (RG) 1.121 criteria for tube structuralintegrity, and (2) leakage in a faulted condition remains below that assumed in calculating the allowable offsite radiation dose limits.

The guidance of Generic Letter (GL) 95-05 was utilized.

The proposed change would preserve the reactor coolant flow margin and reduce the radiation exposure incurred in the process of plugging or repairing the S/G tubes (approximately 0.060 man-rem per tube of exposure would be saved for a plugging -

operation). Other benefits of not plugging TSP indications that meet the alternate plugging criteria (APC) would be a reduction in man-hours and potentialimpact to critical path time during refueling outages.

TVA's goal is to prolong S/G life over the expected remaining plant life. This goal is best achieved by proactive measures that defer or eliminate the need to replace S/Gs. S/G replacement results from the loss of-tube plugging margin.

Accordingly, the proposed TS change would prolong S/G life and reduce personnel l exposure while maintaining the SON S/G plugging margin.

Justification for Chanoes The proposed APC for SON can be summarized as follows:

Tube Support Plate APC Tubes with bobbin indications exceeding the 2.0-volt APC voltage repair limit and less than or equal to 5.4 volts are plugged or repaired if confirmed as flaw indications by a rotating pancake coil (RPC) inspection. Bobbin indications greater than 5.4 volts attributable to ODSCC are repaired or plugged independent of RPC confirmation.

i Operating Leakage Limits j l

Plant shutdown will be implemanted if normal operating leakage exceeds 150 gallons per l day per S/G. I Steam Line Break (SLB) Leakage Criterion Projected end-of-cycle SLB leak rate from tubes left in service, including a probability of detection adjustment and allowances for nondestructive examination (NDE) uncertainties and ODSCC growth rates, must be less than 4.3 gallons per minute for the S/G in the faulted loop. If necessary to satisfy the allowable leakage limit, additionalindications less than the repair limit shall be plugged or repaired.

l Tube Burst Conditional Probability -

The projected end-of-cycle SLB tube burst conditional probability shall be calculated and compared with the value 1 X 102 as defined in GL 95 05.

Exclusion from Tube Plugging Criteria The APC does not apply to TSP intersections having:

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1. Dent signals greater than 5 volts as measured with the bobbin probe.

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< 2. Mixed residuals of sufficient magnitude to cause a 1-volt ODSCC indication (as measured with a bobbin probe) to be missed or evaluated incorrectly.

3. Circumferentialindications.

These indications shall be evaluated to the TS limit of 40 percent depth.

SON's current TS plugging limit of 40 percent throughwall applies throughout the tube length and is based on the tube structuralintegrity for general area wallloss such as pitting or wear. Tube plugging criteria are based upon the conservative assumptions that the tube to TSP crevices are open (negligible crevice deposits or TSP corrosion) and that J the TSPs are displaced under accident conditions. The ODSCC existing within the TSPs is thus assumed to be free-span degradation under accident conditions and the principal requirement for tube plugging considerations is to provide margins against tube burst in accordance with RG 1.121. The open crevice assumption leads to maximum leak rates

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compared with packed crevices and also maximizes the potential for TSP displacements under accident conditions.

One pulled tube with two TSP intersections from SON Unit 1 support ODSCC as the ,

dominant corrosion mechanism consistent with the Electric Power Research Institute (EPRI) database of pulled tubes. The EPRI database, which includes the SON pulled tube data, is more conservative for SLB leak rate analyses and will be used for all SLB analyses.

RG 1.121 guidelines establish the structurallimit as the more limiting of three times normal operating pressure differential (3AP ) or 1.43 times the SLB pressure differential (1.43APst.) at accident conditions. At normal operating conditions, the tube constraint provided by the TSP assures that 3APm burst capability is satisfied. At SLB conditions, the EPRI alternate repair criteria (ARC) are based on free-span indications under the conservative assumption that SLB TSP displacements uncover the ODSCC indications formed within the TSPs at normal operation. From Figure 6-1 of WCAP-13990,the bobbin voltage corresponding to 1.43APste (3,657 pounds per square inch) is 8.82 volts.

The structurallimit is reduced by allowance for NDE uncertainties and crack growth.

The EPRI ARC supplies the NDE uncertainty (WCAP-13990, Section 7.3) at 95 percent uncertainty to obtain an allowance of 20.5 percent of the repair limit. For SON, there is insufficient ODSCC data to define the voltage growth rates, in EPRI Report TR-100407, Draft Revision 1, "PWR Steam Generator Tube Repair Limit - Technical Support  !

' Document for Outside Diameter Stress Corrosion Crack at Tube Support Plates," )

6-the EPRI criteria provides a growth allowance of 35 percent per effective full power l years (EFPY) when plant specific growth data is not available. For SON, the near-term I

cycle lengths are bounded by 1.23 EFPY. The growth allowance for SON is then

! 43.1 percent. The full APC repair limit is obtained by dividing the structurallimit of 8.82 volts by 1.64 (1.0 + 20.5 percent for NDE uncertainties and 43 percent for crack voltage growth). Thus, the full EPRI ARC defined repair limit is obtained as 5.4 volts.

This repair limit conservatively bounds the limit obtained by applying either the EPRI database, as described above, or the NRC database additions described in WCAP-13990, Section 5.1.

In addressing the combined effects of loss-of-coolant accident (LOCA), plus safe shutdown earthquake (SSE) on the S/G component (as required by General Design Criteria 2), it has been determined that tube collapse may occur in the S/Gs at some plants. This is the case as the TSP may become deformed as a result of lateralloads at the wedge supports at the periphery of the plate because of the combined effects of the LOCA rarefaction wave and SSE loadings. Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.

There are two issues associated with S/G tube collapse. First, the collapse of S/G tubing reduces the reactor coolant system flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA, which in turn, may potentially increase peak clad temperature. Second, there is a potential that partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse.

Consequently, since the leak-before-break methodology is applicable to the SON reactor coolant loop piping, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design of the plant. The limiting LOCA event becomes either the accumulator line break or the pressurizer surge line break.

LOCA loads for the primary pipe breaks were used to bound the conditions at SON for

' smaller breaks. The results of the analysis using the larger break inputs show that the LOCA loads were found to be of insufficient magnitude to result in S/G tube collapse or significant deformation. The LOCA, plus SSE tube collapse evaluation performed for another plant with Series 51 S/Gs using bounding input conditions (large-break loadings),

is applicable to SON.

Environmental Impact Evaluation The proposed change does not involve an unreviewed environmental question because operation of SON Unit 1 in accordance with this change would not:

1. Result in a significant increase in any adverse environmentalimpact previously evaluated in the Final Environmental Statement (FES) as modified by NRC's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

2. Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SON that may have a significant environmentalimpact.

a m. . me4- _i4..maam...aJe aeel.6 .a a.+ a S A 4 4 A vs .

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4 ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNIT 1 DOCKET NO. 50-327 (TVA SON-TS-9515, REVISION 1)

TVA COMMITMENT

l TVA COMMITMENT TVA will revise the SON steam generator program by October 25,1995, to include the following requirements:

A. "If alternate plugging criteria (APC) is implemented, the following results, distributions, and evaluations will be submitted to the NRC staff within 90 days of unit restart:

1. The results of metallurgical examinations of tube intersections removed from the unit.

2. End-of-cycle (EOC) voltage distribution - all indications found during the inspection regardless of a rotating pancake coil (RPC) confirmation.

3. Cycle voltage growth rate distribution (i.e., from beginning of cycles to EOC).

4. Voltage distribution for EOC repaired indications - distribution of indications presented in item 1 that were repaired (i.e., plugged or sleeved).

5. Voltage distribution for indications left in service at the beginning of the next operating cycle regardless of RPC confirmation- obtained from items 1 and 3 above.

6. Voltage distribution for indications left in service at the beginning of the next operating cycle that were confirmed by RPC to be crack-like or not RPC inspected, i

7. Nondestructive examination uncertainty distribution used in predicting of the EOC l (for the next cycle of operation) voltage distribution. I l

8. Conditional probability of burst during main steam line break (MSLB) evaluation.

9. Totalleak rate during MSLB evaluation.

B. If the APC is implemented, the actions and administrative controls provided in Enclosure 4 become effective.

C. TVA will implement the probe wear inspection / reinspection requirements of Enclosure 4 for one cycle of operation (Cycle 8)."

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i ENCLOSURE 4 SEQUOYAH NUCLEAR PLANTS (SON'S)

STEAM GENERATOR (S/G) PROGRAM PLAN FOR THE

IMPLEMENTATION OF GENERIC LETTER (GL) 95-05 i

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1. The inspection guidance discussed in Section 3 of Attachment 1 of the GL 95 05 will l be implemented.  !

3.b Rotating pancake coil (RPC) for the purposes of the technical specification (TS)

J change, also includes the use of comparable or improved nondestructive examination techniques, i 3.b.1 TVA will inspect all bobbin flaw indications with voltages greater than lower voltage repair limit volts utilizing a RPC probe.

3.b.2 TVA will inspect all intersections where copper signals interfere with the detection of flaws utilizing a RPC probe. Any indications found at such 4 intersections with RPC should cause the tube to be repaired.

! 3.b.3 TVA willimplement the following dent sample program:

1 Dent Sample Program initiallmolementation of Dent Samolino Plan:

The initial sample in S/Gs 1 and 2 shall be 100 percent of the total hot-leg (HL) dented tube support plate (TSP) population in S/Gs 1 and 2.

! The initial sample for S/Gs 3 and 4 will be 100 percent of the dented TSP intersections at the first and second HL TSPs'and 20 percent of the dented intersections at the third HL TSP.

1 The dent examinations will be performed with a technique qualified to l Appendix H of the Electric Power Research Institute (EPRI) Steam Generator Examination Guidelines. A RPC inspection will be performed. Alternate i probes, such as the Cecco probe, which has demonstrated detection capability for axial and circumferentiat indications comparable to or better than the RPC probes, can be used for these inspections. RPC is used as a general term to reflect an acceptable technique.

. If the RPC inspection of dented intersections identifies circumferential Indication or an axialindication not detected by bobbin, the RPC inspection shall be expanded consistent with Table 1. The result classification as defined in TS Section 4.4.5.2 shall be utilized.

Future Outaae Dent Samole Selection and insoections:

The S/G tube inspection result classification and the corresponding action required shall be as specified in Table 1. The dent inspection frequency shall be performed coinciding with the S/G surveillance requirements. If an unscheduled mid-cycle S/G surveillance is required, the dented TSP inspection shall be performed.

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- The initial sample in S/Gs' 1 and 2 shall be 100 percent of the total HL dented

- TSP population in S/Gs 1 and 2.

1 The initial sample in S/Gs 3 and 4 will be 20 percent of the total HL dented TSP population in S/Gs 3 and 4. )

l The dented TSP intersections selection for S/Gs 3 and 4 will begin at the l

lowest HL TSP elevations, which has the highest probability that stress  !

- corrosion cracking will occur. The initial sample will be 20 percent of the total HL dents in the respective S/G and randomly distributed at the first HL TSP.

Every two S/G inspections,100 percent of the first HL dented TSP intersections will be inspected.

I" If the RPC inspection of dented intersections identifies circumferential indication or an axial indication not detected by bobbin, the RPC inspection shall be expanded consistent with the Table 1. The result classification as defined in TS Section 4.4.5.2 shall be utilized.

Expansion samples would be selected from the lowest HL dented TSP intersections and continue to higher TSP elevations.

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Table 1 : SON Unit i SGs 3 and 4 Expansion of the HL dented TSP Sample 1

i Initial Sample First Expansion Second Expansion j Result Action Required Result Action Required Result Action Required C-1 None N/A N/A N/A N/A

, C2 Inspect an additional C-1 None N/A- N/A 20% sample of TSP intersections in this SG C-2 Inspect an additional 20% C-1 None sample of TSP intersections

- in this SG C-2 Inspect all remaining TSP l intersections in

! this SG j C-3 Inspect all remaining TSP intersections in

i. this SG and a 20% sample in

other SGs C-3 Inspect all remaining TSP N/A N/A intersections in this SG and a 20% sample in other SGs C-3 Inspect all remaining C-1 in None N/A N/A l TSP intersections in other SG this SG and a 20%

sample in other SGs i

C-2 but inspect an additional 20% N/A N/A not C-3 in sample of TSP intersections other SG in other SG

C-3 in Inspect all remaining TSP N/A N/A other SG intersections in other SGs TSP = dented hot-leg tube support plate t

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l 3.b.4 TVA willinspect allintersections with large mixed residuals utilizing a RPC probe. Any indications found at such intersections with RPC should cause the tube to be repaired. j 3.c.1 TVA will use a bobbin coil calibrated against a reference standard used in the )

laboratory as part of the development of the voltage-based approach, through the use of a transfer standard.

3.c.2 TVA will comply with a *10 percent probe variability. TVA willincrease the number of probe samples to 20. Testing will be performed at the mix frequency. TVA plans to follow the industry approach to probe variability that was presented to NRC in November 1994.

Probe wear inspection /reinspections will be governed by the following:

If the last probe-wear standard signal amplitudes prior to probe replacement exceed the f_15 percent limit by a value of X percent, then any indications j measured since the last acceptable probe wear measurement that are within 1 X percent of the plugging limit must be reinspected with the new probe. For j example, if any of the last probe wear signal amplitudes prior to probe  ;

replacement were 17 percent above or below the initial amplitude, then  !

indications that are within 2 percent (17 - 15 percent) of the plugging limit i must be reinspected with the new probe. Alternatively, the voltage criterion may be lowered to compensate for the excess variation, for the case above, amplitudes,> 0.98 times the voltage criterion would be subject to repair.

3.c.4 TVA data analysts will be trained and qualified in the use of analysis guidelines and procedures.

, 3.c.5 Data analysts will use quantitative noise criteria guidelines in the evaluation of 2 the data. However, it is expected that these criteria will be evolving over the inspection and as a result, are subject to change. Data failing to meet these criteria should be rejected, and the tube will be reinspected.

2. If the alternate plugging criteria is implemented, SON will pull a minimum of two tubes and four intersections during the Unit 1 Cycle 7 outage and implement a tube pull program consistent with the GL.

3. In support of Section 4.4.5.4.a.10 of the proposed TS change, the methodology used for calculating the (1) conditional probability of burst, (2) methods for projecting EOC voltage distributions, (3) upper voltage repair limit, and (4) total leak rate during main steam line break, will be in accordance with Westinghouse Electric Corporation, WCAP-14277," Steam Line Break Leak Rate and Tube Burst Probability Analysis Method for ODSCC at TSP Intersections, January 1995."

i The data sets for burst pressure verses bobbin voltage will contain all applicable data consistent with the latest revision of the industry data base as approved by NRC with the latest tube pull data.

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4. SON Abnormal Operating instruction 24 provides instructions on the trending and

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response to rapidly increasing leaks. This instruction is a defense-in-depth measure that provides a means for identifying leaks during operation to enable repair before such leaks result in tube failure.  !

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l ENCLOSURE 5 ADDITIONAL INFORMATION REQUESTED

- DURING THE AUGUST 28,1995, PHONE CALL REGARDING UNIT 1 STEAM GENERATOR (S/G) TUBES i

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1. De it distribution voltage data from tube support locations with primary water stress corrosion cracking identified in Unit 1 Cycle 6.

SG ROW COL LOCATIONS CYCLE CHAN VOLTS 3 1 56 H01 C6 M1 5.4 15 67 H02 C6 M1 13.4 21 64 H01 C6 M1 14.0

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4 3 30 H02 C6 M1 6.8 10 36 H01 C6 M1 24.4 14 25 H01 C6 M1 42.0 21 44 H01 C6 M1 46.3 36 59 H01 C5 M1 5.5 38 35 H01 C6 M1 10.7

2. Representative dent distribution data as of Unit 1 Cycle 6 inspection by voltage and TSP elevation.

j UNIT 1 S/G 3 VOLTS H01 H02 H03 H04 H05 H06 H07 TOTAL

> = 5 < 10 393 367 205 392 50 237 35 1679

> = 10 < 20 237 184 47 286 8 151 10 923

> = 20 < 30 39 26 1 67 1 27 1 162

> = 30 < 40 11 7 0 14 1 4 0 37 I

> = 40 < 50 2 2 0 11 0 2 0 17 l > = 50 2 1 0 7 0 1 0 11 UNIT 1 S/G 4 l

VOLTS H01 H02 H03 H04 H05 H06 H07 TOTAL

! > = 5 < 10 390 338 395 256 67 7 27 1480

> = 10 < 20 461 246 428 214 22 0 6 1377

> = 20 < 30 247 92 189 59 2 0 0 589 l > = 30 < 40 185 28 84 22 0 0 4 323 i > = 40 < 50 166 12 41 12 0 0 0 231

> = 50 233 5 33 2 1 0 2 276 I 3. General Considerations for Accident Condition Analy:qs (Section 4.1 of SON WCAP-13990) l The following information addresses the applicability of analyses performed for the l Farley Nuclear Plant steam generators (S/Gs) for tube deformation under combined l loss-of-coolant accident (LOCA) plus safe shutdown earthquake (SSE) and for tube j stresses under combined SSE plus steam line break / feed line break (SLB/FLB)

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(WCAP-12871) to SON Units 1 and 2. There are severalinputs that were reviewed in order to establish the applicability of the analysis results for Farley to SON Units 1 and 2. These inputs include the tube support plate (TSP) loads resulting from a SSE,

! LOCA induced TSP loads, the deformation characteristics of the TSP under localized in-plate loads, and combined SSE plus SLB/FLB loads. The applicability of the Farley analysis to SON Units 1 and 2 in each of these areas is summarized below.

Aoolicability of Seismic Loads:

The seismic loads for the Farley analysis are taken from a generic seismic analysis for Series 51 S/Gs. The generic analysis is performed using an umbrella spectra that was generated from the plant specific spectra for a number of plants with Series 51 S/Gs.

The plant specific spectra for SON Units 1 and 2 were included in the generation of

! the umbrella spectra. Thus, the TSP loads from the umbrella analysis, which were

! used for the Farley evaluation, are also applicable to the SON units.

Acolicability of LOCA Induced TSP Loads:

Both Farley and SON have been qualified for leak-before-break for their primary

piping. Thus, the limiting LOCA event for both of these plants is a branch line break

(see Attachment A). However, the tube deformation calculations for Farley were performed using TSP loads for the most limiting large break LOCA event. A transient dynamic analysis for Farley for both primary piping and branch line breaks shows the 4

primary breaks to result in TSP loads that are three to four times higher than the branch line breaks. It has subsequently been determined that the induced pressure loadings from a large piping break at Farley will umbrella loadings from a branch line ,

break for SON Units 1 and 2. Thus, using the large pipe break loads for Farley to calculate flow area reduction provides a conservative basis for the SON branch line i breaks.

I Acolicability of TSP Deformation Characteristics:

l The plate deformation characteristics used in the Farley analysis are based on crush j tests performed for Series 51 S/Gs. The plate geometry and wedge configuration (load transfer locations) are the same for both Farley and SON Units 1 and 2. Thus, the plate deformation characteristics are the same for both plants. Since the loads i

used to calculate flow area reduction for Farley are a conservative basis for SON i Units 1 and 2, the flow area reduction calculations will be conservativo.

Aoolicability of Combined SSE Plus SLB/FLB Loads:

l Combined SSE plus SLB/FLB loads were evacuated for Farley relative to the potential for SSE induced bending stress to reduce the burst pressure for the tubes. The effect on burst strength is a function of the SSE bending stresses at TSP locations. Since the seismically induced tube stresses are the result of a generic analysis that umbrellas the SON Units 1 and 2 spectra, the SSE stresses used in the Farley analysis

, also apply to SON. Therefore, the discussion relative to the effect on burst strength of the combined SSE plus SLB/FLB stresses for Farley also applies to SON.

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3-Based on the above, it is concluded that the analyses performed for Farley in WCAP-12871 also apply to the SON Units 1 and 2 S/Gs.

4. Comparison of 0.740-Inch Versus 0.720-inch Diameter Bobbin Probes During the Unit 2 Cycle 6 outage at Sequoyah, Westinghouse demonstrated their 0.720 inch diameter, long life bobbin probe. .This probe displayed a signal-to-noise ratio that was better than the 0.740-inch and 0.720-inch MULC probes used for the .

inspection. Additionally, the long life probe did not exceed the probe wear tolerance as quickly as the MULC probe did, when using the alternate plugging criteria (APC) four-hole standard.

A study was performed that compared the tube data collected with the 0.720-inch long life and the 0.740-inch MULC probes during the last inspection. A total of 116 tubes in S/G 2 were examined with a 0.720-inch long life probe. Thirty-seven tubes were reported as no detectable degradation by both probes. Twenty tubes had 0.720-inch MULC data (no 0.740-inch MULC data) and one had only MRPC data.

This left 58 tubes for comparison. Eighty seven calls were reported in the 58 tubes.

The sample set provides a good cross section of the indications and anomalies seen in Unit 2. The indications include cold-leg thinning, anti vibration bar wear, and primary water stress corrosion cracking below the hot-leg tubesheet. Anomalies include dents and copper deposits. Detectability and sizing were the two areas of concern addressed in the comparison.

l Concerning detectability, three permeability variation (PVN) calls were not reported -

during analysis of the long life probe data. Reviewing the data shows that all were present.

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' Concerning sizing, the amplitude of PVN calls varied the most between the two probe  !

types. This may be due to differences in size, strength, and orientation of the

magnets in the two probes. Also, most PVN signals occur over a short range.

i Analysts may select different locations within this range to report the call. These reasons may explain why differences for these signals ranged from -41 percent to

. + 34 percent. The large variability, coupled with the fact that reduced PVN is a positive quality, led to not consider PVN signals for sizing.

Eliminating the calls made with only one probe and PVN signals leaves 59 tubes to

use for a sizing comparison. Listings for the majority of these 59 signals, along with a summary of the results, can be found in Attachment 1.

Further information was gathered from the October 1990 EPRI report, " Eddy-Current l Probe Characterization." Although many types and sizes of probes were included in the report, the 0.740-inch and 0.720-inch diameter bobbin coils were concentrated l on. During the testing the sensitivity of different probes to volumetric and planar

( tube wall degradation was evaluated. Signal amplitude was the factor used for

measuring the sensitivity. In all tests, the voltages were normalized to a 360-degree dent on the sample.

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No attempt was made to optimize signallevels of each scan by positioning the probe closer to the side of the tube containing the flaws. Test pieces were scanned as in the field using a motorized pusher-puller and a MlZ-18 tester. Some variations existed from multiple scanning of the same flaw using the same probe. The largest change in signal amplitude is caused by adding extension cables. Adding these cables reduced coil impedance, reducing signal amplitude response.

To accomplish the task of flaw detection, especially small volume flaws such as axial cracks, it is important to have an optimum frequency providing the highest flaw detection sensitivity included as one of the four operating frequencies. - EPRI found that for tube wall thickness in the range 0.043 to 0.050 inch, the best detection frequency providing optimal flaw signal responses were centered around 200 -

300 kilohertz (kHz). TVA elected to use 200 kHz as one of their test frequencies prior to the Unit 2 Cycle 6 outage (optimum flaw detection frequency is different for narrow-groove probes).

Conclusions from EPRI's testing show a difference in signal amplitude between 0.740-inch and 0.720-inch bobbin coils, especially when test frequencies exceed 500 kHz. For the range of frequencies used at Sequoyah (10 kHz - 400 kHz),

differences in signal amplitude for the two probe sizes is minimal. In all cases, the 0.720-inch probe had a larger signal amplitude. These results help to support the comparison. Attachment 2 contains figures from the EPRI report showing voltage responses for various types of indications at different frequencies.

In conclusion, it appears that using a 0.720-inch diameter bobbin probe, in lieu of the 0.740-inch probe, will not detract from the eddy current inspections performed at-SON. This study shows that flaws will be detected and changes in amplitude, from

. those called in history with a 0.740-inch probe, should be minimal. Also, using a 0.720-inch probe will eliminate the required probe change when testing begins below

, Row 20. A 0.720-inch probe can be used all the way down to Row 3. If TVA elects to perform an APC exam during the Unit 1 Cycle 7 outage, the Westinghouse long life probe has been shown to stay within tolerance better than the MULC probes used during Unit 2 Cycle 6. Should the long life probe perform as well this outage, savings j in time and dose would occur. Lastly, the fill factor with a 0.720-inch probe is 0.86 (a 0.740-inch probe fill factor is 0.91), which is above the recommended minimal fill factor of 0.80.

5. Comparison of Unit 1 Cycle 6 RPC inspection to WCAP-13990 Appendix A l

Requirements i

RPC Anoendix A Issues

1. Electron discharge machined (EDM) notch standard calls for a simulated support

, plate ring 270 degree 3/4 inch thick.

2. EDM notch standard voltage setup calls for setting 400 kHz and 300 kHz to 20 volts on the 100 percent EDM notch.

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1. EDM notch standard did not have a support ring.

2. EDM notch standard voltage was set at 5.0 volts on the 60 percent indication for )

i 400 kHz. (

This would result in the voltage reading for the 100 percent EDM notch to be 79.37 volts at 400 kHz and 70.82 volts at 300 kHz. Utilizing this setup in Unit 1 Cycle 6 outage would have resulted in voltage reading being four times higher than those if using the setup in Appendix A of WCAP-13990. Since indications were plugged on detection without regard to voltage, there is no significant differences in the data collected and analyzed in Unit 1 Cycle 6 than that which will be collected and analyzed in the Unit 1 Cycle 7 refueling outage.

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ATTACHMENT 1 COMPARISON OF A740MULC & EB720LL PROBES DURING U2C6 i

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l Attachment 1 The following listing is a comparison of 740 and 720 )

<< diameter bobbin coil probes. The 740 probe is an A740MULC l Danufactured by Zetec, while the 720 is Westihghouse's long life probe, EC720LL. All data is from the U2C6 Sequoyah outage.

The comparison will be broken into three sections; quantifiable indications, dents, and other indications and ,

Gnomalies. l a

section 1 - Quantifiable Indications  :

E2w Esl Probe Volts Call Location 3 44 740 1.14 64% HTS-1.03 720 1.24 64%

7 59 740 3.64 93% HTS-1.43 720 4.08 92%

8 53 740 5.36 95% HTS-0.72 720 4.91 94% l 8 57 740 4.46 86% HTS-1.57 720 5.35 83% l 1

8 63 740 1.63 73% HTS-1.88 720 1.65 76%

8 66 740 2.68 70% HTS-0.80 720 2.80 67%

12 92 740 1.38 32% C01 720 1.21 35%

15 19 740 4.78 15% H02+24.4 720 4.14 12%

17 28 740 1.10 30% AV3 720 1.24 33%

19 28 740 0.89 27% AV1 )

720 0.89 28% l l

18 89 740 0.39 20% C01 720 0.35 22%

21 37 740 0.45 61% H01+42.53 720 0.45 56%

22 41 740 0.51 24% AV2 720 0.49 21%

23 69 740 0.96 31% AV2 720 0.87 28%

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EnE .rd21 Probe Dit;g Call Location

23 69 740' O.71 27% AV3 720 0.89 .28%

24 61 740 0.78 29% AV2 720 0.89 28%

24' 63' 740 0.82 29% AV2 720- 0.85 27%

24 67 740 1.81 41% AV2 720 2.05 40%

24 67 740 2.02 42% AV3 720 ~ 2.17 41%

25 54 740 12.21 90% HTE+20.41 720 13.77 90%

26 70 740 0.66 27% AV3 720 1.15 32%

27 65 740, 0.64 26% AV2 720 0.73 26%

27 65 740 0.40 20% AV3 720 0.38 18%

29 32 740 0.55 28% AV2 720 0.87 28%

29 32 740 0.67 31% AV3 720 0.75 26%

29 32 740 0.66 31% AV4 720 0.62 24%

29 37 740 0.32 18% AV2 720 0.45 20%

29 37 740 0.25 16% AV3 ,

720 0.19 12%  !

29 42 740 0.66 27% AV1 720 0.75 26%

29 42 740 0.77 29% AV2 1 720 0.91 29%

'29 42 740 1.41 37% AV3 720 .1.57 36%

~ . - . . . . . . - . . - . . . . . . . . . . . . . - - . ,

EeX sel Probe. Volts Call ' Location AV4

" 29 42 740 -0.31 18%

7 '9 0.30 15%

l 29 56 740' O.34 21%- C04 1 720 0.33 17%

30 44 :740. 0.48 9% H04+2.66 j 720 0.59 11%  ;

1 32- 55 740 1.69 40% AV3 720 1.75 38%

Col l 32 79 740' 1.35 32%

i 720 1.42 41%

33 62 740 0.38 20% AV3 720 0.56' 22%'  !

34 17 740 0.79 15% Col 720 0.68 19%

34- 63 740 0.39 20% AV1 720 0.53 22%

I 1

34 63 740 1.66 39% AV2 720 1.59 36% l 34 63 740 0.49 23% AV3 720 0.57 23%

34 64 740 1.92 41% AV2 720 1.88 39%

36 18 740 1.72 22% C01 720 1.79 23%

37 21 740 0.51 16% C01 720 0.51 14%

38 24 740 1.30 29% C01 720 1.24 20%

42 31 740 2.01 25% C01 l 720 1.58 30%

' 42 64 740 1.35 20% C01  ;

720 1.13 17%

.42' 66 740 0.63 14% C01 '

720 0.61 23%

B.gw .qpl- Probe Volts Call Location

43. 63 740- 0.67 27% Col

720 0.58 26%

, er Indications within the hot leg tubesheet generally had a larger voltage response with the 720 probe, but the measured-porcent throughwall was close for all calls. The majority of

. cold leg thinning calls show a larger voltage with the 740 probe.

- Throughwall percentages are similar for both probes, with the  !

exception of two tubes which had a difference of 9%. In each circumstance one of the signals was slightly distorted, thus -l producing the large difference in throughwall measurements. .

Voltages for AVB calls vary between the two probes (probably due to probe motion caused by the geometry of the u-bend), but the msasurements for the wear signals is similar.

Section 2 - Dents Egw fg;tl Probe Volts Call Location l

3 94 740 2.22 DNT H06-0.85 720 2.43  :

! 5 4 740 6.65 DNT HTS +30.23 .

! 720 5.63 6 1 740 6.19 DNT H01 720 7.62 l 7 2 740 5.14 DNT H01

! 720 6.33 8 2 740 5.37 DNT H01

, 720 5.88 8 3 740 2.53 DNT C07 s

720 2.49 9 3 740 4.82 DNT H01 i 720 5.78 9 4 740 13.32 DNT H01 720 15.92

11 72 740 5.33 DNT H02+8.01 720 5.13 12 19 740 2.12 DNT C07 i 720 2.15 13 9 740 5.21 DNT H04 720 5.73 23 21 740 8.84 DNT HTS +1.99 720 9.20 i

1 R92 2 21 Probe Volts Call Location 23 21 740 8.62 DNT HTS +5.52 720 8.92 23 24- .740 8.53 DNT C05 720 7.07 24 10 740 8.93 DNT HTS +3.10 720 9.62 24 10 740 11.19 DNT HTS +6.78 720 10.99 24 10 740 4.84 DNT HTS +4.24 720 5.26 There were more dent calls which are not listed here (only about half are listed), but a pattern can be seen from those. .

above. The 720 probe provides a larger voltage response on the majority of the dent calls. If a 720 diameter probe is used during the U1C7 outage some dents which measured just under 2.0

-. volts with the 740 probe may become reportable. .

section 3 - other Indications & Anomalies Row 9 21 Probe Volts Call Location 3 44 740 2.02 CUD H01 720 2.13 7 3 740 8.29 CUD H01 720 9.02 10 43 740 2.39 CUD H01 720 2.93 1

16 73 740 14.74 RIC HTS-1.09 720 15.14 23 16 740 3.54 CUD C05 720 4.15 23 57 740 6.41 IRI AV3+13.38 720 6.19 27 84 740 0.23 IRI H01 720 0.28 31 70 740 0.34 IRI H01

< 720 0.44 35 55 740 0.34 RIC HO2 720 0.45

E2w C21 Probe Volts Call Location 37 76 740 0.63 IRI HTS +41.43 720 0.61 e<

43 59- 740 4.36 CUD C03 l 720 4.84 All the above anomalies, with the exception of two freespan IRI calls, had a larger voltage recorded using the 720 long life probe. Most important is the fact that all the calls from all three sections were detected with both probes.

J i

1

4 f f ATTACHMENT 2 RESULTS FROM EPRI'S ,

EDDY-CURRENT PROBE CHARACTERIZATION i

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' SIGNAL NORMALIZE 0 TO DENT RESPONSE f f f f f f f 100 200 300 400 500 600 700 800 FREQUENCY (KHZ) l Figure 1A. S/N ratios from 720 bobbin coil with a 100-foot

, extension cable in a nominal tube with various flaws.

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' SIGNAL NORMALIZED TO DENT RESPONSE U f f f f f f f j 100 200 300 400 500 600 700 800 l FREQUENCY (KHZ)

Figure 1B. S/N ratios from 740 bobbin coil with a 100-foot extension cable in a nominal tube with various flaws.

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j 100 200 300 400 500 600 700 800

' FREQUENCY (KHZ)

Figure 2A. S/N ratios from 720 bobbin coil in a

nominal tube with various flaws.

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' Figure 2B. S/N ratios from 740 bobbin coil in a l nominal tube with flat-bottomed holes. i

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100 200 300 400 500 600 700 800 FREQUENCY (KHZ)

Figure 3A. S/N ratios from 720 bobbin coil in a nominal tube with axial notches.

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NORMALIZED AMPLITUDE AXIAL EDM w 1/4' X 803 i 7

/p,r e - h,*s.% O

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NO EXTENSION CABLES USED SIGNAL NORMALIZED TO DENT RESPONSE

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t I I t t t 100 200 300 400 500 600 700 800 l

FREQUENCY (KHZ)

Figure 3B. S/N ratios from 740 bobbin coil in a nominal tube with axial notches.