Text

Proposed Tech Specs,Revising TS SR & Bases to Incorporate Alternate S/G Tube Plugging Criteria at TSP Intersections
	ML20086S240
Person / Time
Site:	Sequoyah
Issue date:	07/19/1995
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
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References
	NUDOCS 9508010154
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,

I REACTOR COOLANT SYSTEM SURVEILLANCE RE0V!HEMENTS (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 inspection, this shall be recorded and an adjacent tube shall 7"[

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 R193 is not a reactor coolant pressure boundary (tube end up to

"#f'+

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 Inspection 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.

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.

9508010154 950719 PDR ADOCK 05000327 P

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

REACTOR COOLANT SYSTEM SURVElltANCE REOUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.

As used in this Specification:

1.

Imoerfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be con-sidered as imperfections.

2.

Deoradation means a service-induced cracking, wastage, wear or general corrosion occuring or either inside or outside of a tube.

3.

Deoraded Tube means a tube containing imperfections greater 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.

5.

Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

6.

Pluacino limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.

Plugging limit does Rd not apply to that portion of the tube that is not within the pressure boundary of the reactor coolant system (tube end up to y

the start of the tube-to-tubesheet weld)y 4"'

7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8.

Tube Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

9.

Preservice Inspection 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 Ag+

expect.ed to be used during subsequent inservice inspections.

H ut.

Amendment No. 189 SEQUOYAH - UNIT 1 3/4 4-9 October 20, 1994

9 4

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.

The steam generator shall be determined OPERABLE af ter 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 Following each inservice inspection of steam generator tubes, the a.

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 percent 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 R4@

c.

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.

2ksue E

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

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

a.

No PRESSURE BOUNDARY LEAKAGE, rn.., b b.

1 GPM UNIDENTIFIED LEAKAGE, CP" t:t:1 pri :ry-t: ;;;;=d: y 1;;h:g; through cl' st;s, g;n;r;ter; on:

,11 =

,;r :=y thr;;;r==y ;=; t := g;=;r;t =,

d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

]e.

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

I t

f.

1 GPM leakage at a Reacto: Coolant

System pressure of 2235 1 20 psig

! nth:

from any Reactor Coolant System Pressure Isolation Valve specified in l

Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY a.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and.in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any Reactor Coolant System leakage greater than any one of the 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SNUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With any Reactor Coolant System Pressure Isolation Valve leakage c.

greater than the above limit, isolate the high pressure portion of the af fected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, e

or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4 J URVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within y

each of the above limits by:

Rii MAR 261982 SEQUOYAH - UNIT 1 3/4 4-14 Amendment No.12

REACTOR COOLANT SYSTEM 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 i

likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube 150 leakage between the primary coclant system and the secondary coolant system (primary-to-secondary leakage 500$allons 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 the/ loads imposed during ncrmal operation and by postulated accidents./'Oper:ttng plent; h:v demonstrated that primary-to-secondary leakage of-See gallons per day per 5,3,DA steam generator can readily be detected.by radiation monitors of steam 6s Leakage in excess of this limit will require plant generator blowdown, heduled insp_e.ction, during "which the leaking tubes will be shutdown and an unsc located and plugged. k ff G

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

Hve.

secondary coolant.

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

Q'regir l.J Plugging will be required for all tubes with imperfections ex 1 ::11 thickne::.

The portion of the f'g" plugging limit of 40% of the tube nomin: tube that the plugging limit does not apply Sew,.baa is not within the RCS pressure boundary (tube end up to the start of the tube-R191

,Rego % d to-tubesheet weld). The tube end to tube-to-tubesheet weld portion of the tube does not affect structural integrity of the steam generator tubes and therefore

< g, g' g 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 wali thickness.

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

Are 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-R40 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 1

The RCS leakage detection systems required by this specification are 1

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 Systems," May 1973.

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

l I

REACTOR COOLANT SYSTEM BASES f

i 3/4.4.6.2 OPERATIONAL 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 I

low to~ ensure early detection of additional leakage.

The surveillance requirements _ for RCS Pressure Isolation 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.

The 10 GPM IDENTIFIED LEAKAGE lim'if.ation provides allowance for a limited

. amount of leakage from known sources whose presence will not interfere with

the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow l

supplied to the reactor coolant pump seals exceeds 40 GPM with the modulating valve in the supply ifne fully open at a nominal RCS pressure of 2235 psig.

This limitation ensures that'in the event of a LOCA, the safety injection flow I^*[

'll' not be less than assumed in the accident analyses.

{

H m.

/

The tot steam generator be leakage limit f 1 GPM for all s am generators ens es that the dosag contribution fr the tube leakage

11 be 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

}

h the assumptions sed in the analys of these accide s.

The 400 gpd

+

w 1ea e limit per ste generator ensure that steam genera r tube integrity 1

is ma tained in the eve of a main steam ine rupture or un r LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

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

i 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due-to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant' System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may.be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval

. permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentra-tions to within the Steady State Limits.

I SEQUOYAH - UNIT 1 B 3/4 4-4 SEP 17 880 September 17, 1980

{

1 Insert A 4.4.5.2.b.4 Tubes left in service as a result of application of the tube support plate l

alternate plugging criteria shall be inspected by bobbin coil probe during future refueling outages.

l Insert B 4.4.5.2.d.

Implementation of the steam generator tube support plate alternate plugging criteria requires a 100 percent bobbin coil inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length, insert C 4.4.5.4.a.6. This definition does not apply to tube support plate intersections if the

(

l alternate plugging criteria are being applied. Refer to 4.4.5.4.a.10 for the plugging limit applicable to these intersections.

Insert D 4.4.5.4.a.10 Tube Sucoort Plate Pluoaina Limjlis used for the disposition of a steam generator tube for continued service that is experiencing outside diameter j

stress corrosion cracking confined within the thickness of the tube support l

plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

l Degradation attributed to outside diameter stress corrosion cracking a.

within the bounds of the tube support plate with bobbin voltage less than or equal to 2.0 volts will be allowed to remain in service.

b.

Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged except as noted in 4.4.5.4.a.10.c below.

Indications of potential degradation attributed to outside diameter c.

stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to j

5.4 volts may remain in service if a rotating pancake coil inspection l

does not detect degradation. Indication of outside diameter stress corrosion cracking degradation with a bobbin coil voltage greater than 5.4 volts will be plugged or repaired independent of RPC confirmation.

l I

Insert E 4.4.5.5.d For implementation of the alternate plugging criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service should any of the following conditions arise:

1.

If estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing licensing basis assumptions) during the previous operation cycle.

i 2.

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

3.

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

4.

If primary water stress corrosion cracking indications are identified where alternate plugging criteria is being applied.

5.

If the calculated conditional burst probabilit/ based on the actual measured end-of-cycle voltage distribution would have exceeded 1 x 10-2, notify the NRC and provide an assessment of the safety i

significance of the occurrence.

l Insert F f

i 3.4.6.2.c Primary-to-secondary leakage shall be limited to 150 gallons per day through any one steam generator, insert G The repair limit for ODSCC at tube support plate intersections is based on the analysis contained in WCAP-13990 "Sequoyah Unit 1 and 2 Steam Generator Tube Plugging Criteria For Indications At Tube Support Plates, May 1994", and documentation contained in EPRI Report TR-100407, Revision 1, "PWR Steam Generator Tube Repair Limits - Technical Support Document for Outside Diameter Stress Corrosion Cracking at Tube Support Plates." The application of this criteria is based on assuring tubes accepted for continued service wil! retain adequate structural and leakage integrity during normal operating transients and postulated accident conditions such as main steam line break.

Insert H I

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 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 uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 4.3 gpm in the faulted loop, which will limit the calculated offsite doses to within 10% 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 service in order to reduce the postulated primary-to-secondary steam line break leakage to below 4.3 gpm.

l

)

REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued)

All nonplugged tubes that previously had detectable wall pene-1.

trations (greater than 20%).

Tubes in those areas where experience has indicated potential 2.

problems.

A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall 3.

be performed on each selected tube.

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

A Hert-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:

The tubes selected for these samples include the tubes from 1.

those areas of the tube sheet array where tubes with imperfections were previously found.

The inspections include those portions of the tubes where 2.

imperfections were previously found.

}

Tube degradation identified in the portion of the tube that RI.

NOTE:

is not a reactor coolant pressure boundary (tube end up to inseI the start of the tube-to-tubesheet weld) is excluded from B

N * "-

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

One or more tubes, but not more than 1% of the C-2 total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are i

degraded tubes.

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

In all inspections, previously degraded tubes must exhibit Note:

significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

SEQUDYAH - UNIT 2 3/4 4-11 Amsn'dment No.181 October 20. 1992.

~

f i

REACTOR COOLANT SYSTEM t'

\\

SURVEILLANCE REQUIREMENTS (Continued) i

'4.4.5.4 Accentance_ Criteria As used in this Specification:

l a.

Imperfection means an exception to the dimensions, finish or i

contour of a tube from that required by fabrication drawings or 1.

j Eddy-current testing indications below 20% of

~

specifications.

may be con.

the nominal tube wal1~ thickness, if detectable, sidered as imperfections.

i

Dearadation means a service-induced crac' king,

wastage, wear or general corrosion occurring on either inside or outside of a 2.

tube.

i Dearaded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by 3.

degradation.

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

affected or removed by degradation.

i Defect means an imperfection of such severity that it exceeds A tube containing a defect is defective.

i 5.

I the plugging limit.

beyond which i

_Plucoina limit means the imperfection depth at orthe t 6.

Plugging limit does not R2 of the nominal tube wall thickness.

apply to that portion of the tube that is not withi Inserf

)

N the start of the tube-to-tubesheet weld y c

gue l

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-7.

rity 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.

Tube insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the 8.

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

7g g I

DHsre)&

i Amendment No. 181 SEQUDYAH - UNIT 2 3/4 4-13 October 20, 1994

=.

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 9.

Preservice inspection means an inspection of the full length of 1

each tube in each steam generator perf ormed by eddy current techniques prior to service to establish a baseline condition of the tubing.

This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to i

be used auring subsequent inservice inspections.

b.

The steam generator shall be determined OPERABLE af ter 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 Following each inservice inspection of steam generator tubes, the a.

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 the completion of the inspection. This Special Report shall include:

1.

Number and extent of tubes inspected.

2.

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

3.

Identification of tubes plugged, Results of steam generator tube inspections which fall into Category c.

C-3 shall be reported pursuant to Specification 6.6.1 prior to resump-R2 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 i

recurrence.

ITn5M/b 6

L+-

Here November 23, 1984 SEQUOYAH - UNIT 2 3/4 4-14 Amendment No. 28

REACTOR COOLANT SYSTEM m

OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERAT10ti 3.4.5.2 Reactor Coolant System leakage shall be limited to:

No PRESSURE BOUNDARY LEAFAGE, a.

b.

1 GPM UNIDENTIFIED LEAKAGE,

' CP" tet:1 pr e:ry-t:-::cend:ry 1::k:g: through al' st::: g:ncr: tere

^ #

i c.

g j,,,

nd 500 gal'en ;;r d:y three;5 :ny en: Ste:r ger:rct:r.-

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

40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of e.

2235 2 20 psig, 2235 ! 20 psig 1 GPM leakage at a Reactor Coolant System pressure of fece any Reactor Coolant System Pressure Isolation Valve specified in f.

Table 3.4-1.

APPLICABILITY: HODES 1, 2, 3 and 4 ACTION:

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

a.

With any Reactor Coolant System leakage greater than any one of the b.

above limits, excluding PRES 5URE BOUNDARY LEAKAGE, and leakage f rom l

Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

A SEQUOYAH - UNIT 2 3/4 4-18

REACTOR COOLANT SYSTEM l

i L

-BASES-i 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes

-ensure that the structural integrity of this portion of'the RCS will be main '

tained. ~The program for-inservice inspection of steam generator tubes is Inservice based on a modification of Regulatory Guide 1.83, Revision 1.

inspection of steam generator tubing is essential'in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or. inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube-degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the 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 The extent of cracking during likely result in stress corrosion. cracking.

plant operation would be limited by the limitation of. steam generator tube fp leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage

-50Fga11ons per day per st.eam generator).

5%=M Cracks having a primary-to-secondary leakage less than this limit during hs operation will have an adequate margin of safety.to withstand the loads imposed during normal operation and by postulated accident,s._.f Operatir.; phnts

.have demonstrated that primary-to-secondary leakage of 40Fgallons per day per steam generator can readily be detected by radiation monitors of steam 9enerator blowdo Leakage in excess of this limit will require plant shutdown and an un heduled insJection rin which the leaking tubes will be ocated and plugged.

or b ps o ff-p Wastage-type defects are unlikely with proper chemistry treatment of the

(;

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

Here.

h

_ p. = Plugging will be required for all tubes with imperfections exceeding t eThe portion of the

S:1 ::P thiebe::.

9e Q p1 ;;f ng,i:!t cf " cf the tub:

tube that the plugging limit does not apply to is the portion of the tube that d6dh is not within the RCS pressure boundary (tube end up to the start of the tube-s,4%en.

to-tubesheet weld).

The tube end to tube-to-tubesheet weld portion of the g1 e:0..rmt tube does not affect ~ structural integrity of the steam generator tubes and h therefore 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.

Lsart Whenever the results of any steam generator tubing inservice inspection

I H

fall. into Category C-3, these.results will be promptly reported to the Commis-Hart sion pursuant' to Specification 6.9.1 prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in' a requirement for analysis, laboratory examinations, tests, addition-al eddy-current inspection, and revision of the Technical Specifications, if necessary.

SEQUOYAH - UNIT 2 B 3/4 4-3 Amendment No. 181

REACTOR COOLANT SYSTEM BASES

)

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 provided to monitor and detect leakage from the Reactor Coolant Pre Boundary.

Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL 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 Isolation Valves provide added assurances of valve integrity thereby reducing the probability of gross Leakage from the RCS isolation valve failure and consequent intersystem LOCA.

l valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with 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 exr.eeds 40 GPM with the modulating ht valve in the supply line fully open at a nominal RCS pressure of 2235 psig.

i 1

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

Here-Thet)alsteamgenerato tube leakage limit of 1 G) for all s eam generators n t isolated from t RCS ensures that the dos e contribu ion from he tube leaka e will be limite to a small f ction of Par 100 limit in the e nt of either s' team generator tube rupture r steam line break.

The 1 GPM li 't is consiste t with the assum ions used in the analysis f these ac idents.

The 0 gpd leakag limit per steam enerator ens es that ste generator tube 1 tegrity is a 'ntained in the e nt of.a main \\ steam line pture or under L A conditions.

\\

t PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may Therefore, be indicative of an impending gross f ailure of the pressure boundary.

the presence of any PRESSURE BOUNDARY LEAXAGE requires the unit to be promptly

, placed in COLD SHUTDOWN.

SEQUOYAH - UNIT 2 B 3/4 4-4 1

insert A 4.4.5.2.b.4 Tubes left in service as a result of application of the tube support plate alternate plugging criteria shall be inspected by bobbin coil probe during future refueling outages.

i Insert B j

4.4.5.2.d.

Implementation of the steam generator tube support plate alternate l

plugging criteria requires a 100 percent bobbin coil inspection for hot-leg -

tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

{

Insert C 4.4.5.4.a.6. This definition does not apply to tube support plate intersections if the i

alternate plugging criteria are being applied. Refer to 4.4.5.4.a.10 for t.he plugging limit applicable to these intersections.

l Insert D i

4.4.5.4.a.10 Tube Suooort Plate Pluaaina Limit is used for the disposition of a steam generator tube for continued service that is experiencing outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on l'

maintaining steam generator tube serviceability as described below:

Degradation attributed to outside diar.ieter stress corrosion cracking i

a.

within the bounds of the tube support plate with bobbin voltage less than or equal to 2.0 volts will be allowed to remain in service.

l b.

Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged except as noted in 4.4.5.4.a.10.c below.

1 c.

Indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate l

with a bobbin voltage greater than 2.0 volts but less than or equal to 5.4 volts may remain in service if a rotating pancake coil inspection does not detect degradation. Indication of outside diameter stress corrosion cracking degradation with a bobbin coil voltage greater than 1

5.4 volts will be plugged or repaired independent of RPC confirmation.

1 I

I l

f.

~.

Insert E 4.4.5.5.d For implementation of the alternate plugging criteria to tube support plate intersections, notify the Commission prior to returning the steam 1

generators to service should any of the following conditions arise:

1.

If estimated leakage based on the actual measured end-of cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing licensing basis assumptions) during the previous operation cycle.

1 2.

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

3.

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

4.

If primary water stress corrosion cracking indications are identified where alternate plugging criteria is being applied.

5.

If the calculated conditional burst probability based on the actual measured end-of-cycle voltage distribution would have exceeded 1 x 102, notify the NRC and provide an assessment of the safety significance of the occurrence.

Insert F l

3.4.6.2.c Primary to-secondary leakage shall be limited to 150 gallons per day l

through any one steam generator.

l Insert G The repair limit for ODSCC at tube support plate intersections is based on the analysis contained in WCAP-13990 "Sequoyah Unit 1 and 2 Steam Generator Tube Plugging Criteria For Indications At Tube Support Plates, May 1994", and documentation contained in EPRI Report TR-100407, Revision 1, "PWR Steam Generator Tube Repair Limits - Technical Support Document for Outside Diameter Stress Corrosion Cracking at Tube Support Plates." The application of this criteria is based on assuring tubes accepted for continued service will retain adequate structural and leakage integrity during normal operating transients and postulated accident conditions such as main steam line break.

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.

I

Insert I The total steam generator tube leakage limit of 600 gallons per day for all steam 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 uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 4.3 gpm in the faulted loop, which willlimit the calculated offsite doses to within 10% 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 f aulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 4.3 gpm.

. _.. _... ~..

i ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION (TS) Cl4ANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-15)

DESCRIPTION AND JUSTIFICATION FOR TS AMENDMENT j

l

4 Descriotion of Chanae TVA proposes to modify the SON Units 1 and 2 technical specifications (TSs) to incorporate new requirements associated with steam generator (S/G) tube inspection and repair. The new requirements establish attemate S/G tube plugging criteria at tube support plate (TSP) intersections. The proposed changes are as follows:

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

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

2. Add SR 4.4.5.2.d.

" Implementation of the steam generator tube support plate alternate plugging criteria requires a 100 percent bobbin coil inspection for hot-leg tube support plate intersections and cold leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indicatiors.

The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length."

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

"This definition does not apply to tube support plate intersections if the alternate plugging criteria are being applied. Refer to 4.4.5.4.a.10 for the plugging limit l

applicable to these intersections."

4. Add SR 4.4.5.4.a.10

" Tube Sucoort Plate Pluaaina Limit is used for the disposition of a steam generator tube for continued service that is experiencing outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

Degradation attributed to outside diameter stress corrosion cracking within the a.

bounds of the tubes support plate with bobbin voltage less than or equal to 2.0 volts will be allowed to remain in service.

b.

Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged except as noted in 4.4.5.4.a.10.c below.

. c.

Indications of potential degradction attributed to outside diameter stress corrosion cracking within the bounds of the tube support pla:3 with a bobbin voltage greater than 2.0 volts but less than or equal to 5.4 volts may remain in service if a rotating pancake coil inspection does not detect degradation.

Indication of outside diameter stress corrosion cracking degradation with a bobbin coil voltage greater than 5.4 volts will be plugged or repaired independent of RPC results."

i

5. Add SR 4.4.5.5.d.

I "For implementation of the alternate plugging criteria to tube support plate intersections, notify the staff prior to returning the steam generators to service I

should any of the following conditions arise:

1.

If estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing licensing basis assumptions) during the previous operation cycle.

2.

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

3.

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

4.

If primary water stress corrosion cracking indications are identified where alternate plugging criteria is being applied.

i 5.

If the calculated conditional burst probability based on the actual measured end-of-cycle voltage distribution would have exceeded 1 x 102, notify NRC and provide an assessment of the safety significance of the occurance.

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 operational leakage limits.

l l

i i

l

1 i Reason for Chanae TVA is proposing to change SON Unit 1 and 2 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 intersectior.s 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 off site radiation dose limits. The guidance of draft Generic Letter (GL) 94-XX 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 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 goalis 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 exposure while maintaining the SON S/G plugging margin.

Justification for Chanaes The proposed alternate plugging criteria for SON can be summarized as follows:

Tube Support Plate Alternate Plugging Criteria (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 RPC inspection. Bobbin indications greater than 5.4 volts attributable to ODSCC are repaired or plugged independent of RPC confirmation.

Operating Leakage Limits Plant shutdown will be implemented if normal operating leakage exceeds 150 gallons per day per S/G.

Steam Line Break (SLB) Leakage Criterion Projected end-of-cycle SLB leak rate from tubes left in servica, including a probability of detection (POD) adjustment and allowances for ncndestructive 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, additional indications less than the repair limit shall be plugged or repaired.

t

4 Tube Burst Conditional Probability The projected end-of-cycle SLB tube burst conditional probability shall be calculated and compared with the value 1 X 10-2 as defined in the draft GL 94-XX.

Exclusion from Tut's Plugging Criteria The APC does not arply to TSP intersections having:

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

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

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 wall loss 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 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 agairst tube burst in accordance with RG 1.121. The open crevice assumption leads to maximum Sak rates 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 EPRI database of pulled tubes. The EPRI database, which includes the SON pulled tube data, is more conservative for SLB leak rate analyses using draft NUREG-1477 methodology than the data obtained from the SON pulled tubes. Therefore, the more conservative EPRI database is used for all SLB analyses.

RG 1.121 guidelines establish the structurallimit as the more limiting of three times normal operating pressure differential (3APuo) or 1.43 times the SLB pressure differential (1.43APsa) at accident conditions. At normal operating conditions, the tube constraint provided by the TSP assures that 3APuo burst capability is satisfied. At SLB conditions, the EPRI alternatt repair criteria (ARC) are based on free-span indications under the conservative asramption 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 1.43APsw (3,657 pounds per square inch [psil is 8.82 volts).

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

l The EPRI ARC supplies the NDE uncertainty (WCAP-13990, Section 7.3) at 95 percent j

uncertainty to obtain an allowance of 20.5 percent of the repair limit. For SON, there is insufficient prior ODSCC data to define the voltage growth rates. In EPRI Report l

i TR-100407, Draft Revision 1, "PWH Steam Generator Tube Repair Limit - Technical Support Document for Outside Diameter Stress Corrosion Crack at Tube Support Plates,"

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 alternate plugging criteria repair limit is obtained by dividing the j

i structurallimit of 8.82 volts by 1.64 (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 j

the EPRI database, as described above, or the NRC database additions described in j

WCAP-13990, Section 5.1.

l in addressing the combined effects of loss-of coolant accident (LOCA), plus safe f

shutdown earthquake (SSE) on the S/G component (as required by GDC 2), it has been determined that tube collapse may occur in the S/Gs at some plants. This is the case as a

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 l

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 RCS 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 i

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 :oads 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.

l

-. l i

Environmental Impact Evaluation i

The proposed change does not involve an unreviewed environmental question because i

operation of SON Unit 1 in accordance with this change would not:

i

1. Result in a significant increase in any adverse environmentalimpact previously i

evaluated in the Final Environmental Statement (FES) as modified by NRC's testimony i

to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

i

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.

i P

c l

i I

6 E

l t

t i

9 F

t i

1

.. m

i I

1 i

f ENCLOSURE 3 i

PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-15)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION i

f 5

l i

f

4 Significant Hazards Evaluation TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) in accordance with the proposed amendment will not:

1.

Involve a significant increase in the probability or consequences of an accident previously evaluated.

Testing of model boiler specimens for free-span tubing (no tube support plate restraint) at room temperature conditions shows burst pressures in excess of 5,000 pounds per square inch (psi) for indications of outer diameter stress corrosion cracking with voltage measurements as high as 19 volts. Burst testing performed on intersections pulled from SON with up to a 1.9-volt indication shows measured burst pressure in excess of 6,600 psi at room temperature. Burst testing performed on pulled tubes from other plants with up to 7.5-volt indications shows burst pressures in excess of 5,200 psi at room temperatures. Correcting for the effects of temperature on material properties and minimum strength levels (as the burst testing was done at room temperature), tube burst capability significantly exceeds the safety-factor requirements of NRC Regulatory Guide (RG) 1.121.

Tube burst criteria are inherently satisfied during normal operating conditions because of the proximity of the tube support plate (TSP). Since tube-to-tube support plate proximity precludes tube burst during normal operating conditions, use l

of the criteria must retain tube integrity characteristics that maintain a margin of safety of 1.43 times the bounding faulted condition steam line break (SLB) pressure differential. During a postulated SLB, the TSP has the potential to deflect during blowdown following a main SLB, thereby uncovering the TSP intersections.

Based on the existing database, the RG 1.121 criterion requiring maintenance of a safety factor of 1.43 times the SLB pressure differential on tube burst is satisfied by 7/8-inch-diameter tubing with bobbin coil indications with signal amplitudes less than 8.82 volts (WCAP-13990), regardless of the indicated depth measurement. A 2.0-volt plugging criterion (resulting in a projected end-of-cycle (EOCl voltage) compares favorably with the 8.82 volt structura! limit considering the extremely slow apparent voltage growth rates and few numbers of indications at SON. Using the established methodology of RG 1.121, the structurallimit is reduced by allowances for uncertainty and growth to develop a beginning of cycle (BOC) repair limit that would preclude indications at EOC conditions that exceed the structural limit. The nondestructive examination (NDE) uncertainty component is 20.5 percent, and is based on the Electric Power Research Institute (EPRI) alternate repair criteria (ARC).

l l

. Test data indicates that tube burst cannot occur within the TSP, even for tubes that have 100 percent throughwall electro-discharge machining notches,0.75 inch long, provided that the TSP is adjacent to the notched area. Because of the few number of indications at SON, the EPRI methodology of applying a growth component of 35 percent per effective full power year (EFPY) will be used. Near-term operating cycles at SON are expected to be bounded by 1.23 years, therefore, a 43 percent growth component is appropriate. When these allowances are added to the BOC alternate plugging criteria (APC) of 2.0 volts in a deterministic bounding EOC voltage of approximately 3.26 volts for Cycle 7, operation can be established. A 5.56-volt deterministic safety margin exists (8.82 structurallimit - 3.26-volt EOC equal 5.56-volt margin).

For the voltage / burst correlation, the EOC structurallimit is supported by a voltage of 8.82 volts. Using this structurallimit of 8.82 volts, a BOC maximum allowable repair limit can be established using the guidance of RG 1.121. The BOC maximum allowable repair limit should not permit the existence of EOC indications that exceed the 8.82-volt structurallimit. By adding NDE uncertainty allowances and an allowance for crack growth to the repair limit, the structurallimit can be validated.

Therefore, the maximum allowable BOC repair limit (RL) based on the structurallimit of 8.82 volts can be represented by the expressions:

RL + (0.205 x RL) + (0.43 x RL) = 8.82 volts, or, l

the maximum allowable BOC repair limit can be expressed as, j

l RL = 8.82-volt structurallimit/1.64 = 5.4 volts.

This RL (5.4 volts) is the appropriate limit for APC implementation to repair bobbin indications greater than 2.0 volts independent of rotating pancake coil (RPC) confirmation of the indication. This 5.4-volt upper limit for nonconfirmed RPC calls is consistent with other recently approved APC programs (Farley Nuclear Plant, Unit 2).

The conservatism of the growth allowance used to develop the repair limit is shown f

by the most recent SON eddy current data. Two tubes plugged in Unit 1 during the last outage had less than one volt of growth over the past five operating cycles.

Only seven tubes in Unit 2 required repair because of outside diameter stress corrosion cracking (ODSCC) at the TSP intersections.

Relative to the expected leakage during accidient condition loadings, it has been previously established that a postulated ma'n SLB outside of containment, but upstream of the main steam isolation valve (MSIV), represents the most limiting radiological condition relative to the APC. Implementation of the APC will determine t

l whether the distribution of cracking indications at the TSP intersections is projected to be such that primary-to-secondary leakage would result in site boundary doses within a small fraction of the 10 CFR 100 guidelines. A separate analysis has

. determined this allowable SLB leakage limit to be 4.3 gtllons per minute (gpm) in the faulted loop. This limit uses the TS reactor coolant systam (RCS) lodine-131 activity level of 1.0 microcuries per gram dose equivalent lodin.s-131 and the recommended lodine-131 transient spiking values consistent with NLREG-0800. The analysis method is WCAP-14277, which is consistent with the guidance of the NRC draft generic letter (GL) and will be used to calculate EOC leakage. Because of the relatively low number of indications at SON, it is expected that the actualleakage values will be far less than this limit. Additionally, the current lodine-131 levels at SON range from about 25 to 100 times less than the TS limit.

Application of the criteria requires the projection of postulated SLB leakage, based on the projected EOC voltage distribution for Cycle 8 operation. Projected EOC voltage distribution is developed using the most recent EOC eddy current results and a voltage measurement uncertainty. Data indicates that a threshold voltage of 2.8 volts would result in throughwall cracks long enough to leak at SLB condition.

The draft GL requires that allindications to which the APC are applied must be included in the leakage projection. Tube pull results from another plant with 7/8-inch tubing with a substantial voltage growth database have shown that tube wall degradation of greater than 40 percent throughwall was readily detectable i

either by the bobbin or RPC probe. The tube with maximum throughwall penetration of 56 percent (42 average) had a voltage of 2.02 volts. The SON Unit 1 pulled tube had a 1.93-volt indication with a maximum depth of 91 percent and did not leak at SLB condition. Based on the SON pulled tube and industry pulled tube data supporting a lower threshold for SLB leakage of 2.8 volts, inclusion of all APC intersections in the leakage modelis quite conservative. The ODSCC occurring at SON is in its earliest stages of development. The conservative bounding growth estimations to be applied to the expected small number of indications for the upcoming inspection should result in very smalllevels of predicted SLB leakage.

Historically, SON has not identified ODSCC as a contributor to operational leakage, in order to assess the sensitivity of an indication's BOC voltage to EOC leakage potential, a Monte Carlo simulation was performed for a 2.0-volt BOC indication.

The maximum EOC voltage (at 99.8 percent cumulative probability) was found to be 4.8 volts. The leakage component from an indication of this magnitude, using either the NUREG-1477 or EPRI leakage models, is 0.12 or 0.028 gpm, respectively.

Therefore, as implementation of the 2.0-volt APC does not adversely affect steam generator (S/G) tube integrity and implementation will be shown to result in acceptable dose consequences, the proposed amendment does not result in significant increase in the probability or consequences of an accident previously evaluated.

L 2.

Create the possibility of a new or different kind of accident from any previously

analyzed, implementation of the proposed S/G tube APC does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism that could result in an accident outside of the region of the TSP elevations; no ODSCC is occurring outside the thickness of the TSP. Neither a single or multiple l

tube rupture event would be expected in a S/G in which the plugging criteria is applied (during all plant conditions).

TVA will implement a maximum leakage rate limit of 150 gallon per day per S/G to help preclude the potential for excessive leakage during all plant conditions. The i

SON TS limits on primary-to-secondary leakage at operating conditions include a maximum of 0.42 gpm (600 gallons per day [gpd)) for all S/Gs, or, a maximum of 150 gpd for any one S/G. The RG 1.121 criterion for establishing operational leakage rate limits that require plant shutdown is based upon leak-before-break considerations to detect a free-span crack before potential tube rupture during faulted plant conditions. The 150-gpd limit should provide for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack l

resulting in leakage that is associated with the longest permissible crack length.

RG 1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded. The longest permissible crack is the length that provides a factor of safety of 1.43 against bursting at faulted conditions maximum pressure differential. A voltage amplitude of 8.82 volts j

I for typical ODSCC corresponds to meeting this tube burst requirement at a lower 95 percent prediction limit on the burst correlation coupled with 95/95 lower tolerance limit material properties. Alternate crack morphologies can correspond to 8.82 volts so that a unique crack length is not defined by the burst pressure versus voltage correlation. Consequently, typical burst pressure versus through-wall crack length correlations are used below to define the " longest permissible crack" for evaluating operating leakage limits.

The single through-wall crack lengths that result in tube burst at 1.43 times the SLB pressure differential and the SLB pressure differential alone are approximately 0.57 inch and 0.84 inch, respectively. A leak rate of 150 gpd will provide for detection of 0.4-inch-long cracks at nominal leak rates and 0.6-inch-long cracks at i

the lower 95 percent confidence level leak rates. Since tube burst is precluded during normal operation because of the proximity of the TSP to the tube and the potential exists for the crevice to become uncovered during SLB conditions, the leakage from the maximum permissible crack must preclude tube burst at SLB l

conditions. Thus, the 150-gpd limit provides for plant shutdown before reaching l

critical crack lengths for SLB conditions. Additionally, this leak-before-break i

evaluation assumes that the entire crevice area is uncovered during blowdown.

f Partial uncover will provide benefit to the burst capacity of the intersection.

l I.:

5 As S/G tube integrity upon implementation of the 2.0-volt APC continues to be maintained through in-service inspection and primary-to-secondary leakage monitoring, the possibility of a new or different kind of accident from any accident previously evaluated is not created.

3.

Involve a significant reduction in a margin of safety.

The use of the voltage based APC at SON is demonstrated to maintain S/G tube integrity commensurate with the criteria of RG 1.121. RG 1.121 describes a method acceptable to the NRC Staff for meeting General Design Criteria (GDC) 14,15,31, and 32 by reducing the probability or the consequences of S/G tube rupture. This is accomplished by determining the limiting conditions of degradation of S/G tubing, as established by in-service inspection, for which tubes with unacceptable cracking should be removed from service. Upon implementation of the criteria, even under the worst-case conditions, the r,ccurrence of ODSCC at the TSP elevations is not expected to lead to a S/G tube rupture event during normal or faulted plant conditions. The EOC distribution of crack indications at the TSP elevations will be confirmed to result in acceptchlo primary-to-secondary leakage during all plant conditions and radiological consequences are not adversely impacted.

In addressing the combined effects of loss-of-coolant accident (LOCA), plus safe f

shutdown earthquake (SSE) on the S/G component (as required by GDC 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 RCS 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 (PCT). 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. Therefore, at SON, no tubes will be excluded from using the voltage repair criteria due to deformation of collapse of S/G tubes following a LOCA plus an SSE.

6-Addressing RG 1.83 considerations, implementation of the bobbin probe voltage based interim tube plugging criter.a of 2.0 volt is supplemented by: (1) enhanced eddy current inspection guidelines to provide consistency in voltage normalization, (2) a 100 percent eddy current in.ipection sample size at the TSP elevations, and (3) RPC inspection requirements for the larger indications left in service to characterize the principal degravation as ODSCC.

As noted previously, implementation of the TSP elevation plugging criteria will decrease the number of tubes that must be repaired. The installation of S/G tube l

plugs reduces the RCS flow margin. Thus, implementation of the alternate plugging criteria will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.

Based on the above, it is concluded that the proposed license amendment request does not result in a significant reduction in margin of safety.

1 l

G 0

ENCLOSURE 4 -

PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-15)

TVA COMMITMENT l

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

A.

"If alternate plugging criteria 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 (a) 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 (a) and (c) 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.

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

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

9. Totalleak rate during MSLB evaluation."

B.

"If the alternate plugging criteria is implemented, the administrative controls provided in Enclosure 5 become effective."

e a

ENCLOSURE 5.

RESPONSES / EXCEPTIONS TO DRAFT GENERIC LETTER (GL) 94-XX GUIDANCE l

l l

i

.'d TVA will implement the requested actions for each of the eight requested actions of the f

i draft GL with the following comments / exceptions:

t 1.

The inspection guidance discussed in Section 3 of Enclosure 1 of the draft GL will be implemented with the following comments / exceptions:

f 3.b Rotating pancake coil (RPC) for the purposes of the TS change, also i

includes the use of comparable or improved nondestructive examination

}

techniques.

3.b.1 TVA will inspect all bobbin flaw indications with voltages greater than 2.0 volts utilizing a RPC probe.

j 3.b'.3 TVA will inspect all intersections where copper signals interfere with the.

detection of flaws utilizing a RPC probe, f

3.b.4 TVA will inspect all dents greater than 5 volts at tube support plate elevations with known ODSCC.

4 3.b.5 TVA will inspect all intersections with large mixed residuals utilizing a RPC i

probe.

3.c.3 Due to time constraints for the fall'95 Unit 1 outage, new probes certified to a 10 percent variability are not available, it is anticipated that probes meeting the variability requirement will be available within six months of the final GL being issued and will be used when available and as necessary.

j 3.c.4 The requirement to reinspect all tubes if the wear measurement exceeds 15 percent is unnecessary. As acknowledged in the draft GL, a 5.6-volt repair criterion is justified; however, the repair criterion is limited to T

2.0 volts. To require reinspection of all tubes inspected with a specific bobbin probe if probe wear reaches 16 percent is not necessary from a safety standpoint and could affect critical path outage time.

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

If the last probe-wear-standard signal amplitudes pior to probe replacement exceed the f_15 percent limit by a value of X percent, then any indications measured since the last acceptable probe wear measurement that are within X percent of the plugging limit must be i

i reinspected with the new probe. For 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 must be re-inspected with the new probe. Alternatively, the voltage criterion may be lowered to compensate for the excess variation, for the case above, amplitudes 2. 0.98 times the voltage criterion would be subject to repair.

O e 3.c.6 Data analysts will use quantitative noise criteria guidelines in the i

evaluation of the data. However, it is expected that these criteria will be evolving over the inspection and as a result, are subject to change.

Inspections will be performed in accordance with Appendix A in WCAP-13990, "Sequoyah Unit 1 and 2 Steam Generator Tube Plugging Criteria For Indications At Tube Support Plates," May 1994, as further detailed in WCAP-14277.

2.

Calculations of the leakage will be in accordance with the guidance of Section 2.b of of the draft GL with the following responses / exceptions:

2.b The calculations performed in support of the voltage based repair criteria will follow the methodology described in WCAP-13990, "Sequoyah Unit 1 and 2 Steam Generator Tube Plugging Criteria For Indications At Tube Support Plates," May 1994.

2.b.2(1) No distribution cutoff will be applied to the voltage measurement variability distribution.

2.b.3(1)/

2.b.3(2) TVA understands that the NRC Staff has approved Reference 1, Criteria 2a and 2b and concurred with all data excluded under Criteria 1b and 1c. Data will not be excluded under Criteria 3a,3b, or 3c unless approved by the NRC Staff.

2.b.4 in order to preclude the possible need for rapid turn around of a technical specification amendment, the reactor coolant system specific iodine activity will not be revised.

3.

Calculation of the conditional burst probability will be per the guidance of Section 2.a of Enclosure 1 of the draft GL with following responses / exceptions:

2.a The calculations performed in support of the voltage-based repair criteria will follow the methodology described in WCAP-13990,"Sequoyah Unit 1 and 2 Steam Generator Tube Plugging Criteria For Indications At Tube Support Plates," May 1994, as further detailed in WCAP-14277.

2.a.1 TVA understands Reference 1 data exclusion under Criteria 2a and 2b has been approved by the NRC Staff.

4.

The operationalleakage limits for SON will be changed to 150 gallons per day.

5.

SON Abnormal Operating Instruction 24 provides instructions on the trending and response to rapidly increasing leaks.

[

l

~

4 j 4

6.

The tube pull guidance of Enclosure 1, Section 4 of the draft GL will be followed.

i 7.

The results will be reported in accordance with Enclosure 1, Section 6 of the draft GL.

8.

The paragraph associated with mid-cycle inspection limits has been deleted pending issuance of the final GL and revised repair limit formulas.

.sf in addition to the eight items listed above, TVA provides the following information in response to Section 1.b.1 of the draft GL.

Analysis performed for SON indicates no tubes will be excluded from using the voltage repair criteria due to deformation of collapse of steam generator tubes following a loss of coolant accident plus a safe shutdown earthquake event.

V e

t I

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ML20086S240

Text

Dearadation means a service-induced crac' king,

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