ML20128H978

Proposed TS Modifying Insp Requirements of TS 3/4.4.5 Re SGs & LCO for 3/4.4.6.2 Re Operational Leakage to Establish Interim Plugging Criteria

ML20128H978
Person / Time
Site:	Summer
Issue date:	02/12/1993
From:	SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML20128H971	List: ML20128H978 RC-93-0039, Application for Amend to License NPF-12,modifying Insp Requirements in TS 3/4.4.5, SGs & LCO for 3/4.4.6.2, Operational Leakage to Avoid Unnecessary Plugging of SG Tubes W/O Compromising Safety
References
NUDOCS 9302170258
Download: ML20128H978 (61)

Text

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ATTACHMENT 1 THE TECHNICAL SPECIFICATION PAGES AS PROPOSED BY - .,

THIS AMENDMENT REQUEST p$g21Jgggg,ogggoj,3 p PDR- ,

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, . - Attichm:nt 1.to Docum2nt Control Lctt:r .

TSP 920001-1 Page 1 of 1 LIST OF AFFECTED PAGES Page- Specification Description of Change 3/4'4 3/4.4.5 Corrected reference in SR 4.4.5.2 from 4.4.5.2.d to 4.4.5.2.e.

3/4 4-12 3/4.4.5 Corrected reference in 4.4.5.2.b.3 and added new

" item e" to SR 4.4.5.2.

3/4 4-13 3/4.4.5 Repagination 3/4 4-14 3/4.4.5 Clarified that ODSCC will not apply to existing plugging limit of.SR-4.4.5.4.a. item 6.

3/4 4-14a 3/4.4.5 Defined scope of inspection related to use of TSP-IPC in SR 4.4.5.4.a. item:9.

3/4 4-15 3/4.4.5 Added a new " item 17" to SR 4.4.5.4.a.

3/4 4-15a 3/4.4.5 Continued new " item 17,"

-revised " plugging limits" to " applicable limits" in SR 4.4.5.4.b and added TSP-IPC results.to SR 4.4.5.5.d.

3/4 4-19 3/4.4.6.2 Modified the LCO to~ ,

decrease the allowed SG leakage from 1 gpm to-.31 gpm total and from 500 gpd

.to 150 gpd in one SG.

B 3/4 4-3 Bases 3/4.4.5 _ Updated. bases'to reflect changes in specifi_ cation (150-gpdleakageper generator limit and basis for ODSCC plugging limit)._-

B 3/4 4-4 Bases-3/4.4.6.2 Incorporated new-lower-leakage limits-into bases discussion.

e REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. '

EPLICABILITY: MODES 1, 2, 3 and 4.

ACT'ON:

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

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of-the following augmented inservice inspection program and the requi:ements of Specification 4.0.5.

4.4.5,1 Steam Generator Samole Selection and Insoection - Each steam generator shall be determineo OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generatorsispecified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selectio'n and Inspection - The-steam generator tune minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam' generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall' be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators.. The tube inspections will also-include at least 3% of the total number of sleeved tubes-in all 3 steam genera-tors or all of the-sleeved tubes in the generator chosen for the inspection program, whichever is less. These inspections will include both the. tube a the sleeve. Whenever applying the exceptions of 4.4.5.2.a through 4.4.5.2 E previous defects or imperfections in the area of the steam generator tube-repaired by sleeving are not considered an area requiring reinspection. The tubes selected for these inspections shall-be' selected on a random basis except:

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

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

SUMMER - UNIT 1 3/4 4-11 Amendment No. 59'

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1.-

All nonplugged tubes that previously had detectable. wall penetrations greater than 20% that were not repaired.

2.

Tubes in tb-se areas where experience has indicated potential problems.

3. A tube inspection (pursuant to Specification 4.4.5.4.a.

be performed on each selected tube. If any selected tube does shall not permit the passage of the eddy current probe for a tube inspection, this shall be recorded.and an adjacent tube shall be select:d and subjected to a tube inspection.

c.

In addition to the sample required in 4.4.5.2 b.1 through 3, all tubes which have had the F* or L* criteria applied will be inspected in the tueesheet region. These tubes may be excluded from 4.4.5.2 b.1 I provided the only previous wall penetration of >20% was located below the F* distance or the required L* inspection area (3.5 inches). I d.

The tubes selected as the second and third samples (if required by ,

Table 4.4-2) during each inservice inspection may be subjected to a '

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 imperfections were previously found.

2. The inspections include those portions of the tubes where

(-[M/r#t. /1, imperfections were previously found.

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

Category 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_ defcctive, 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 l tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit L significant (greater than 10%) further wall punetrations l

to be included in the above percentage calculations.

! SUMMER - UNIT 1 3/4 4-12 Amendment No. 64. 59, 96 '

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INSERT A

e. Implementation-of the interim steam generator tube / tube support plate elevation plugging limit is only applicable for Cycle 8 and-requires a 100% bobbin probe inspection forfall hot leg tube support plate intersections and all cold leg intersections down to the.

lowest cold leg tube support plate with previous outer diameter stresscorrosioncracking(00 SCC) indications. An inspection using the rotating pancake coil (RPC) probe is required in order to assess.

operability of tubes with flaw like bobbin coil' signal amplitudes greater than 1.0 volt but less than 2.2. volts.-For. tubes that will be administrative 1y plugged, no RPC inspection is required. The RPC results are to be evaluated against the criteria contained in WCAP-13522 to determine if plugging is required.

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. s REACTOR COOLANT-SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria

a. As used in this Specification:

1.

Imperfection 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 considered as imperfections.

2.

Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

3.

Degraded Tube means a tube containing imperfections greater tnan or equal to 20% of the nominal wall thickness caused by degradation.

4.  % Degradation means the percentage of the tube wall thickness I affected or removed by degradation. '

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

6.

Tube Plugging or Repair Limit means the imperfection depth at or beyona which the tube shall be repaired (i.e. sleeving) or removed from service by plugging pnd is equal to 40% of the the portiontube nominal. of thewall tubethickness in t (Wtubesheet below tnc " or laf Th E ^ distance provided the tube is not degraded (i.e., no indications

{,g $) of cracks) within the F* distance for F* tubes and within the L* distance for la tubes.

M7 Sleeve Plugging or Repair Limit

a. For the area in the upper weld joint, any degradatiun shall be plugged unless it can be clearly demonstrated by a quali-fied NDE technique that the degradation is less than 40% of the nominal wall thickness of the sleeve for-ID imperfec-tions or-less for 0.0. than 40% nominal wall thickness of the tube imperfections.

b.

For the area of the tube behind the sleeve and above the upper weld joint, tubes with any degradation shall be plugged unless it can be clearly demonstrated by a qualified NDE technique, thst the degradation is less than 40% of the nominal ~ wall thickness.

c. For the area below the upper weld joint, any defect greater than 40% of the nominal sleeve wall thickness shall be plugged.

SUMMER - UNIT 1 3/4 4-14 Amendment No. 54, 59, 92, 96

, 1, INSERT B Also, this= definition does not apply for tubes experiencing outer ,

diameter stress-corrosion cracking confirmed.by bobbin probe inspection- '

to be within the thickness of the tube support plates. See

'4.4.5.4.a.17_for the TSP-IPC plugging limit.

4 INSERT C The scope of the TSP-IPC_ inspection will consist offall the hot leg tube support plate intersections and all cold leg intersections down to and including, at least, the lowest cold leg tube support plate with previous outer diameter stress corrosion cracking (00 SCC) indications, s

M +

e REACTOR COOLANT SYSTDI.

$UtytiLLANCE tt00!#Detkil (Continued)

, 4.

contatna inte a defect large enough to affect its l lossgrity in the event of an Operating lasts tart breat as specif fed in 4.4.5.3.c. above.

9. Tube

( Inse r-f' 3 !nseection noens an inspection of the steam generator tut ie 0free Gne point of entry (het leg side) completely around th g

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3 -bend to the tap support of the cold leg. e-10.

$1eeve the tuM. Inseection seens an inspection of the sleeved portio tube strectly above the upThis inspection'will include 3 Inches j of t

- the new pressure beuneary,per weld, the upper weld which forms upper weld. and the sleeve motorial below the 11.

Romaind tube establishes means a tube that has uneergone a process that its servicoatility. -

will be used when sleeving a steam generater tune:One er more o The used Combustion Engineering per report CDt 337 P. Inc. weld sleeve process will be -

The report tabcock 8AW 2048 & P. Wticex Kinetic sleeve process will be used per 4

o stagga . UNIT 1 3/4 4 14e A8'**t N'* A8 'I 4 - _ _ .

+

i REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued)

12. Preservice Inspection means an inspection of the full length of each tube in each steam generator performud by eddy current >

techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION >

using the equipment and techniques expected to be used during subsequent inservice inspections.

13.

Fa Distance is the oistance into the tubesheet from the face of the tubesheet or the top of the last hardroll, whichever is lower (further into the tubesheet) that has been conservatively chosen to be 1.6 inches.

14. F* TUBE is the tube with degradation, below the F* distance, equal t'o or greater than 40%, and not degraded (i.e. , no indica- '

tions of cracking) within the F* distance.

15.

L* Distance is the distance into the tubesheet from the face of the tubesheet or the top of the last hardroll, whichever is lower (further into the tubesheet), that has been conservatively chosen to be 0.7 inches.

16. La Tube is a tube with short (less than 0.5 inches) axially oriented (20 degrees or less from axial) degradation occurring below %e undegraded La distance. An additional minimum of 1.0 inches of sound expanded tube (below the La distance) separated by no more than 2 areas of axially oriented' degradations must be contained in the top 3.5 inches of tube (within the tubesheet).

Each area of degradation is limited to a maximum of 5 distinct 7s indications. A maximum of 2500 tube ends per steam generator Tubes qualifying as f* tubes are not classified f ut't,p) r mayasutilize L* tubes.L*.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceading the plugging-limit-) required by Table 4.4-2.

iywna t umt3 4.4.5.5 Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2.

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:

SUMMER - UNIT 1- 3/4 4-15 Amendment No. M , M ,

M, U, H, 96

4 4 INSERT D 17.TheTubeSupportPlateInterimPlugging(TSP-IPC)Limitisusedtoassess

- OPERADILITY of a steam generator tube that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates. For application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude of flaw like indications. The plant specific guidelines will accommodate the additional Information needed to evaluste tube support plate signals with respect to the voltage / depth parameters as specified in Specification 4.4.5.2. Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in the V.C. Summer Unit 1 steam generator inspections for consistent voltage normalization.

1. A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.0 volt, regardless of the depth of tube wall penetration, provided that, the projected end of' cycle distribution of crack indications is verified to result in primary to secondary leakage less than 1 gpm under SLB conditions. The basis for determining expected leak rates from the projected crack distribution is provided in WCAP-13522.

2. A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than 2.2 volts provided a rotating pancakecoil(RPC)inspectiondoesnotidentifydefects. "

3. Flaw like indications with a bobbin coil signal amplitude greater than 2.2 volts will be plugged.

Certain tube support plate intersections, as identified in WCAP-13522, will be excluded from application of the TSP-IPC limit due to the potential tube deformation from tube support plate loading during a postulated LOCA plus SSE event. This is to preclude the potential for secondary to primary in-leakage in the tubes during reflood following a postulated LOCA Event.

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!!Ac?OR 000Lut_ systof .

$URVillLANCI #f00]RtM(Kf5 ("ontinued) 1 Number and extent of tubes inspected. ,

2.

. Location and todication of an percent imperfection. of well thickness penetration for each 3.

Identification of tubes plugged or repaired.

c.

Results of stees generator tube inspections which fall into be reported pursuant to 10 CFR $0.72(b)2 plant operation.

submitted to provide a description of investigation determine cause of the tube degradation and corrective measure to prevent recurrence.

d.

Commisssuits of inspections of F* and La tubes shall be report of the unit toa report to the Director, ONRR, prior t he inspection. This r estart

[l 1.

11 inclueet t Identification of F" a nd '

2.

i Locatio re of the degradation I

approval of this report is not rweired orier to restart.

_1 d.

to the Comission in a report to the-Director. ON restart of the unit following the inspection:

1.

Inspection of F* and L* tubes. The re identification of the F* and L* tubes, port shall include size of degradation. and the location and-2.

Inspections of all tubes in which the TSP-IPC limit has been apslied.

p tu)es, and the locationThe report shall include a listing of applicable-and extent of degradation (applicable (voltage). intersections per tube)

NRC approval of this report is not required prior to restart.

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t $UWER - UNIT 1 3/4 4-15e Amendment No, M

REACTOR C00LAN? SYSTEM OPERATIONAL LEAKAGE .

t!MITING CON 0!T10N FOR OPERATION t

Reactor Coolant System leakage shall be limited to:

3.4.6.2

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKACE,

c. 1 GPM total primary-to-secondary leakage through all steam generators -not--4solated-fam-the-ReactoFCoolant-Gystem and

[Te- ._) 500 gallons per day through any one-4 team generator nob-4solated-Q{'"[ h. gyN-ff om-the-ReactoFCoolant-System, f {)

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

e. 33 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig,

f. 1 GPM leakage at a Reactor Coolant System pressure of 2235 + 20 psig from any Reactor Coolant System Pressure Isolation valve specified in Table 3.4-1.

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

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

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

above limits, excluding PRESSURE BOUNDARY LEAKAGE and Leakage from 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 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 SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

c. With any Reactor Coolant System Pressure Isolation Valve Leakage 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 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2,1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by;

a. Monitoring the reactor building atmosphere (gaseous or particulate) radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SUMMER - UNIT 1 3/4 4-19

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INSERT E  ;

0.31 GPM total primary-to-secondary leakage through all steam generators and 150 gallons per riay through any one steam generator when utilizing the 1 acceptance criteria of SR 4.4.5.4.a.17. ,

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REACTOR C00l ANT SYSTEM

1. ASES 3/4.4.5 STEA4 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 based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam genarator

% tubing also provides a means of characterizing the nature and cause of any tube 3 .js n degradation so that corrective measures can be taken.

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.The plant is expected to be operated in a manner such that the seconda 4 coolan 11 be maintained within those chemistry limits found to resu negligible rosion of the steam generator tubes. If the secondaJy-chemistry is no lant intained within these limits, localized co,rreslon may likely result in str Ncor,rosion cracking. The extent pkcFacking during i plantoperationwouldbeTt(tedbythelimitationoksteamgeneratortube leakage between the primary co'af ,

(primary-to-secondary leakage = 50 system s-per and ,the' daysecondary per steam coolant generator). system Cracks _ having a primary to secondaryJe g operation will have an adequate rqargin ofa s (ftless than this the o withstand limitloadsduring imposedduringnormaloperatJon'andbypostulated nts. Operating plants have demonstrated that piniary-to-secondary leakage of allons per day per steam generator cay 1fadily be detected by radiation monitor steam generator blowdown. Le3kage in excess of this limit will require plant shu n and an unscheduled 4nspection, during which the leaking tubes will be locate plugged'or repaired.

=:= =:= = = == =

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. A"#}'  % _aiti_ge secondar t9pe'de'ects Sol nt. However, even are if a unil defectG y=with should develop proper in service, chemistry treatm 6, ' will be found (during4Qeduled inservice steam generator tube-eximinations.

Plugging or repairing wilt'tre-ratquired for all tubes-wit'fi imperfections exceed-ing 40% of the tube nominal wall TMrkneestiam generator tube inspections of operating plants have demonst;tated"the c5pabi QtJ to reliably detect wastage-type degradation thatys-penetrated 20% of the ori~ght I thickness.

Plu gginfMot required for tubes meeting either the FA or Dtter,it.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3. these results will be promptly reported to the Commission pursuant to 10 CFR 5').72(b)2(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 ex6minations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

SUMMER - UNIT 1 B 3/4 4-3

! AmendmentNo.3NENEN 96

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INSER1 F 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 corros!on may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the allowable steam generator primary to secondary tube leakage as specified by the Limiting Condition for Operation 3.4.6.2 " Operational Leakage." Cracks having a primary-to-secondary leakage within these limits during operation will have an adequate  ;

margin of safety to withstand the loads imposed during both normal operation and postuisted accidents. Operating plants have demonstrated that primary-to-secondary leakage of as low as 160 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired.

INSERT G Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, steam generator tube inspections of operating plants have demonstrated the capability to reliably-detect wastage-type degradation that has penetrated 20% of the original tube wall thickness. Plugging or repairing will be required for tubes with degradations, other than outer diameter stress corrosion cracking, that exceed 40% of the tube nominal wall thickness.

Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired based on the criteria of 4.4.5.4.a.17. Also, plugging is not required for tubes meeting either the F* or L* criteria.

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REACTOR COOLANT SYSTEM I

i 8ASES

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

The RCS leakage detection systems required by this specification are provided to monitor and detect leakage free the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of l

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

l 3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has showe that while a limited amount of leakage is expected from the RCS, the unMentified portion of this leakage can be reduced F to a threshold value of less than 1 GPH. This. threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM 10ENTIFIED LEAKAGE limitation provides allowance for a limited-amount of leakage from known sources whose presence will not interfere with the detection of UNIDENT!FIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 33 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 will not be less than assumed in the accident analysas, a

The surveillance requirements for RC5 Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross-valve failure and consequent intersystea ^t0CA. Leakage from the:RCS Pressure _ 4 Isolation valves is IDENTIFIED LEAKAGE and will be considered ~as-a portion of-the allowed limit. '

"Nhe,_ total t stIsm generator tube leakage limit o[1 GPM r al te h" N j generators not4 salated from the RCS ensures that-the dos e ution from

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GN the tube leakage wiiT'tre411tL i ted to a small f ra event of either a steam generhtuhe cuptu(nierbfPart:

limit is consistent with the assumptfo'ns7 sed 100 limits in the or' steam line break. . The 1-GPM the analysis of these.

accidents. The 500ydAsaEige -limit per steam gen - ensures that steam generator tuhmygrity is maintained in the ev' Ant of a sa ta g line ruptyrs- W 'under LOCA conditions.

N SUMER - UNIT l' B 3/4 4a4 ,

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INSERT H t

The total steam generator tube leakage limit.of 1 GPM for all steam generators entures that the-dosage contribution from the tube leakage will be limited to a small fraction of Pert 100 limits in the event of either a steam generator tube rupture or a steam line break. The 0.31 GPM limit for all steam generators associated with implementations of SR 4.4.5.4.a.17 is conservative with respect to the 1 GPM assumed in the analysis of these accidents. The 500 GPD leakage limit per steam generator ensures that steam generator tube integrity will be maintained in the event of a main steam line  !

rupture or under LOCA conditions. The 150 GPD leakage limit associated with implementations of SR 4.4.5.4.a.17 is conservative with respect to the 500 .

GPD limit and meets the RG 1.121 requirement for allowable leakage of the  ;

largest permissible crack.

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REACTOR COOLANT SYST[M 3/4.4.5 STEMi GfNERATORS LJMITIR[,03DJITJ0N FORJPE RATION 3.4.5

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Each steam generator shall be OPERABLE.

APPLICABILJI11 MODES 1, 2, 3 and 4.

eCTION: t With one or more steam generators inoperable, restore the Inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200*f, SURVEILLANCf REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator __Samp_1_e_ Selection and Insp.qcilgn - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and_Jnspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in'all steam generators. The tube inspections will also include at least 3% of the total number of sleeved tubes in all 3 steam genera-tors or all of the sleeved tubes in the generator chosen for the inspection program, whichever is less. These inspections will include both the tube and the sleeve. Whenever applying the exceptions of 4.4.5.2.a through 4.4.5.2.e, l previous defects or imperfections in the area of the steam-generator tube repaired by sleeving are not considered an area requiring reinspection. The tubes selected for these inspections shall be selected on a random basis except;

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

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

SUMMER - UNIT 1 3/4 4-11 AmendmentNo.[jl

REAQTOR COO (ANT SYSTEM ,

SURVEll_ LANCE REQUIREMfHTS (Continued) _

1. All nonplugged tubes that previously had cetectable wall penetrations greater than 20% that were not repaired.

2. Tubes in those areas where experience has indicated potential '

problems.

3. A tube inspection (pursuant to Specification 4.4.5.4.a.9) shall l be performed on each selected tube. If any selected tube does ,

not permit the passage of the eddy current probe for 1 tube inspection, this shall be recorded and an adjacert tube shall be selected and subjected to a tube inspection,

c. In addition to the sample required in 4.4.5.2 b.1 through 3..all tubes which have had the F* or L* criteria applied will be inspected in the tubesheet region. These tubes may be excluded from 4.4.5.2 b.1 provided the only previous wall penetration of >20% was located below the F* distance or the required L* inspection area-(3.5 inches).

d. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a 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 imperfections were previously found.

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

e. Implementation of the interim steam generator tube / tube support plate '

elevation plugging limit is only applicable for Cycle 8 and requires a 100% bobbin probe inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with previous outer diameter stress corrosion cracking (0DSCC) indications. An inspection using the rotating pancake coil (RPC) probe is required in order to assess operability of tubes with flaw like bobbin coil signal amplitudes greater than 1.0 volt but less than 2.2 volts. For tubes that will be administratively plugged, no RPC inspection is required. The RPC results are to be evaluated against the criterion contained in WCAP-13522 to determine if plugging is required.

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

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

SUMMER - UNIT 1 3/4 4-12 AmendmentNo.//, /)l, //

REACTOR COOLANT SYSTEM SURVElttANCE REQUIREMENTS (Continued)

Cateco,ty InspectdonResults 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.

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

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

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-l' category'or if two consecutive inspections demonstrate that previously observed degra-

~

dation has not continued and no additional-degradation has occurred, the inspection interval may be extended to a maximum of once per-40 months.

b. If the results of the inservice inspection of a steam generator-conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.at the interval may then be extended to a maximum of once per 40 months.

c. Additional, unscheduled inservice inspections shall be performed on-each steam generator in accordance with the first sample inspection-specified in Table 4.4-2 during the shutdown subsequent-to any of the following conditions:

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

2. A seismic occurence greater than the Operating Basis' Earthquake.

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

4. A main steam line or feedwater line break.

SUMMER < UNIT 1 3/4 4-13 Amendment No.

-REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREP NTS (Continued) 4.4.5.4 A ceptance Criteri4 J

a. As used in this Specification:

1. Imperfjtdign means an exccption to the dimensions, finish or contour of a tube from that required by fabrication drawings'or specifications. Eddy-current testing indications below 20t of the nominal tube wall thickness, if detectable, mcy be considered as imperfections.

2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either the inside or outside of a tube.

3. Degraded Tube means a tube containing imperfection:, greater than or equal to 20% of the nominal wall thickness caused by degradation.

4.  % Degradation 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 or repair limit. A tube containing a defect is defective.

6. Tube Plu_gqing or Repair Limit means the imperfection depth at or beyond which the tube shall be repaired (i.e. sleeving) or removed from service by plugging and is equal to 40% of the nominal tube wall thickness.

This definition does not apply to the portion of the tube in the tubesheet below the F* or L* distance provided the tube is not degraded (i.e., no indications of cracks) within the F* distance for F* tubes and within the L* distance for L* tubes. Also, this definition does not apply for tubes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates. See 4.4.5.4.a.17 for the TSP-IPC plugging limit.

7. Sleeve Pluggina or Repair limit

a. For the area in the upper weld joint, any degradation shall be plugged unless it can be clearly demonstrated by a quali-fied NDE technique that the degradation is less than 40% of the nominal wall thickness of the sleeve for ID imperfec-l tions or less than 40% nominal wall thickness of the tube for-l 0.0, imperfections.

l l

b. For the area of the tube behind the sleeve and above the upper weld joint, tubes with any degradation shall be plugged unless it can be clearly demonstrated by a qualified NDE technique, that the degradation is-less than 40% of the nominal wall thickness.

SUMMER - UNIT 1 3/4 4-14 Amendment No. ,/,

jgACTOR COOLANT SYSTEM I

SURVEltl ANCE REQUIREMENTS (Continued)

c. For the area below the upper weld joint, any defect greater  ;

than 40% of the nominal sleeve wall thickness shall be ,

plugged.

8. 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.

9. Tube _ Inspection 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. The scope of the TSP-!PC inspection will consist of all the hot leg tube support plate intersections and all cold leg intersections down to and including, at least, the lowest cold leg tube support plate with previous outer diameter stress corrosion cracking (00 SCC) indications.

10. Sleeve Inip_ect_ioj) means an inspection of the sleeved portion of the tube. This inspection will include 3 inches of the parent tube directly above the upper weld, the upper weld which forms the new pressure boundary, and the sleeve material below the upper weld.

11. Repaired tube means a tube that has undergone a process that re-establishes its serviceability. One or more of the following will be used when sleeving a steam generator tube:

The Combustion Engineering Inc. weld sleeve process will be used per report CEN-337-P.

The Babcock & Wilcox Kinetic sleeve process will be used per report BAW-2045 P.

12. Preservice-In_spection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

13. F* Distance is the distance into the tubesheet from the face of the tubesheet or the top of the last hardroll, whichever is-lower (further into the tubesheet) that has been conservatively chosen to be 1.6 inches.

14. F* TUBE is the tube with degradation, below the F* distance, equal to or greater than 40%, and not degraded (i.e., no indica-tions of cracking) within the F* distance.

SUMMER - UNIT 1 3/4 4-14a Amendment No. )(6,///,

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REACTOR COOLANT SYS_TJM_

SURVElllANCE REQUIREMENTS (Continued)

15. l* Dis 1Ancg is the distance into the tubesheet from the face of the tubesheet or the top of the last hardroll, whichever is lower (further into the tubesheet), that has been conservatively chosen tu be 0.7 inches.

16. L* Tu.bg is a tube with short (less than 0.5 inches) axially oriented (20 degrees or less from axial) degradation occurring below the undegraded L* distance. An additional minimum of 1.0 inches of sound expanded tube (below the L* distance) separated by no more than 2 areas of axially oriented degradations must be contained in the top 3.5 inches of tube (within the tubesheet).

Each area of degradation is limited to a maximum of 5 distinct indications. A maximum of 2500 tube ends per steam generator may utilize L*. Tubes qualifying as F* tubes are not classified as L* tubes,

17. The Tube Support Plate Interim Plugging Criteria (ISP-IPC) Limit is used to assess OPERABILITY of a steam generator tube that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates, for application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude of flaw like indications. The plant specific guidelines will accommodate the additional information needed to evaluate tube support plate signals with respect to the voltage / depth parameters as specified in Specification 4.4.5.2.

Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in the V.C. Summer steam generator inspections for consistent voltage normalization.

1. A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.0 volt, regardless of the depth of tube wall penetration, provided that, the projected end~

of cycle distribution of crack indications is verified to result in primary to secondary leakage less than 1 gpm under SLB conditions. The basis for determining expected leak rates from the projected crack distribution is provided in WCAP-13522.

2. A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than 2,2 volts provided a rotating pancake coil (RPC) inspection does not identify defects.

3. Flaw like indications with a bobbin coil signal amplitude greater .

than 2.2 volts will be plugged.

SUMMER - UNIT 1 3/4 4-15 Amendment No. ,$

ps, p .

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• ..-n--e. - - m w.w

d 4 RCACTOR COOLANT SYSTEM

$URVEftLANCE REQUIREMENTS (Continued)

Certain tube support plate intersections, as identified in WCAP-13522, will be excluded from application of the 1SP-IPC limit due to the potential tube deformation from tube support plate loading during a postulated LOCA plus SSE event. This is to preclude the potential for secondary to primary in-leakage in the tubes during reflood following a postulated LOCA Event.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the applicable limits) required by Table 4.4-2. l 4.4.5.5 Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2.

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-tnickness penetration for each indication of an imperfection.

3. Identification of tubes plugged or repaired.

c. Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to 10 CFR 50.72(b)2(i) prior to resumption of plant operation. A report pursuant to_10 CFR 50.73(a)2(ii) shall be submitted to provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken '

to prevent recurrence.

d. Inspection results for the following inspections shall be reported to the Commission in a report to the Director, ONRR, prior to the restart of the unit following_the inspection: i

1. Inspection of F* and L* tubes. The report shall include identification of the f* and L* tubes, and the location and size of the degradation.

2. Inspections of all tubes in which the TSP-IPC limit has been applied. The report shall include a listing of applicable tubes, and the location (applicable intersections per tube) and extent of '

thedegradation(voltage).

NRC approval of this report is not required prior to restart.

SUMMER - UNIT 1 3/4 4-15a AmendmentNo./fl 4

- -.,wc w --, - - - - - - - e ,-- ., - - , . . - . - - - . . . ~ . , - - . , a

i REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE f LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited tot

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UN10ENTiflED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through all steam generators and $00 gallons per day through any one steam generator, or-  :

0.31 GPM total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator when utilizing the acceptance criteria of SR 4.4.5.4.a.17.

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

e. 33 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig.

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

APPLICABILITY: MODES 1, 2, 3 and 4 ACil0N:

a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STAN0BY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and Leakage from 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 STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System Pressure Isolation Valve Leakage greater than the above limi' 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 use of at least two closed manual or deactivated automatic valves, or be in a at least HOT STANDBY within thenext 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .

SURVEILLANCE REQUIREMENTS , 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by;

a. _ Monitoring the reactor building atmosphere (gaseous or particulate)-

radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.  ;

SUMMER - UNIT 1 -3/4 4-19 Amendment

REA 10R COO _LANT SYSim t

BASES l 3/4 4 125 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 based  !

on a modification of Regulatory Guide 1.83, Revision 1. Inservice Inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in toe event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that icad 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 seconriary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the allowable steam generator primary to secondary tube leakage as specified by tne Limiting Condition for Operation 3.4.6.2 " Operational teakage." Cracks having a primary-to-secondary leakage within these limits during operation will have an adequate margin of safety to withstand the loads imposed during both normal operation and postulated accidents.

Operating plants have demonstrated that primary-to-secondary leakage of as low as 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage-type degradation that has penetrated 20% of the c'iginal tube wall thickness. Plugging or repairing will be required for tubes with degradations, other than outer diameter stress corrosion cracking, that exceed 40% of the tube nominal wall thickness. Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired based on the criterion of 4.4.5.4.a.17. Also, plugging is not required for tubes meeting either the F* or L* criteria.

Whenever the results of any steam gererator tubing inservice inspection-fall into Category C-3, these results will be promptly reported to the Commission pursuant to 10 CFR 50.72(b)2(i) prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

SUMMER - UNIT 1 B 3/4 4-3 Amendment No. /Jf, f7, 8e W

REACTOR COOLANnSYSTEM l

BASES 3/4.4.6 _ RJACTOR COOL ANT 51 STEM LEAKAG{

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are 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.

]L h426.2 OPERATIONAL LEAKAg 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 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 sedis exceeds 33 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 will not be less than assumed in the accident analyses. l The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

The total steam generator tube leakage limit of.1 GPM for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or a steam line break. The 0.31 GPM limit for all steam generators-associated with implementations of SR 4.4.5.4.a.17 is conservative with respect to the 1 GPM assumed in the analysis of these accidents. The 500 GPD leakage limit per steam generator ensures that steam generator tube integrity will be maintained in the event of a main steam line rupture or under LOCA conditions. The 150 GPO leakage-limit associated with implementations of SR 4.4.5.4.a.17 is conservative with respect to the 500 GPO limit and meets the RG 1.121 requirement for allowable i leakage of the largest permissible crack.

SUMMER - UNIT 1 B 3/4 4-4 Amendment No.

i i

1 1

ATTACHMENT 2 1 o

i DESCIllPTION OF AMENDMENT REQUEST AND SAFETY  !

EVALUATION

-- - -. 4 - , , - . - - - - -- ., .-,- -

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. . Attcchment 2 to Document Control D:sk 1.etter TSP 920001 1  !

Pcga 1 of 20

• Description of Amendment Itequest CurrenLLjcense Condition - Technical Specification (TS) 3/4.4.5, "Stc.am Generators," states that each steam generator shall be operable in modes 1 through 4. This TS provides further guidance relative to surveillance intervals, steam generator tube sample selection and inspection, and acceptance and reporting criteria. TS 3/4.4.6.2, " Operational Leakege " ,

provides operational leakage limits for modes 1 through 4. Specifically, the.

limiting condition for operation (LCO 3.4.6.2.c) allows a 1 GPM total primary to secondary leakage through all ' steam generators not isolated from the reactor coolant system and 500 gallons per day through any one steam generator not isolated from the reactor coolant system.

tunctio_n of 1he Affectad TechniLal Specificatig_ns - The TS surveillance requirements for inspections of the steam generator tubes relative to TS 3/4.4.5 ensures that the structural integrity of the tubes Will be '

maintained. The program for inservice inspection of the steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

The function of TS 3/4.4.6.2, " Operational Leakage," is to provide a total steam generator leakage limit of 1 GPM for all steam generators not isolated from the RCS. This ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 500 gpd leakage limit per steam generator ensures tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

DescriptionofProposedChanae-Asrequiredby10CFR50.91(a)(1),an analysis is provided to demonstrate that the proposed license amendment to implementaninterimtubepluggingcriteriaforthetubesupportplate(TSP) ,

elevation outside diameter stress corrosion cracking (ODSCC) occurring in the '

V. C. Summer steam generators involves no significant hazards consideration.

The proposed plugging criteria involves correlations between eddy current-bobbin probe signal amplitude (voltage) and tube burst and leakage characteristics. The plugging criteria is based on testing of laboratory induced ODSCC specimens, extensive examination of pulled tubes from operating steamgenerators(industrywide),andfieldexperiencefromleakagedueto indicationsatthetubesupportplatesresultingfromaxialODSCC(identified inEuropeanplantsonly).

The proposed amendment would modify Technical Specifications 3/4.4.5, " Steam-Generators," 3/4.4.6, " Reactor Coolant System Leakage," and the associated bases which provide tube inspection requirements and acceptance criteria to determine the level of degradation-for which a tube experiencing ODSCC at the tube support plate elevations may remain in service in the V. C. Summer steam generators.

For Cycle 8 operation of V. C. Summer, an interim tube support plate elevation plugging criteria is proposed which utilizes a conservative voltage based plugging limit. The interim criteria can be summarized as follows:

!l

Attcchment 2 to Docum:nt Control D:sk Letter TSP 920001 1 P:ge 2 of 20 ,

Flaw indications with a bobbin coil voltage less than or equal to 1.0 volt can remain in service without further action, for flaw indications in excess of 1.0 volt but less than 2.2 volts (preliminaryrepairlimitforfullimpicmentationofthe alternate plugging criteria), the tube can remain in service provided an RPC inspection of the indication does not detect a defect. Indications involving other modes of tubo degradation will be assessed for operability based on the current 40% depth based. F*, and L* plugging criteria. Crack indications above  :

?.2 volts will be plugged or repaired by sleeving regardless of RPCresults(ifRPCtested).

To remain consistent with the V. C. Summer licensing basis, potential leakage following a postulated steam line break event at the end of cycle (EOC) conditions shall be less than 1.0 gpm for the most limiting steam generator.

Off-site doses will consequently remain within a small fraction of the 10CFR100 limits. Bobbin coil flaw indications inspected by RPC and found to have no RPC indications do not need to be included within the leak M e analyses. Bobbin indications less than 1.0 volt, if not inspected by RPC, are to be included in the leakage analyses. If the projected leakage' exceeds 1.0g>mduringapostulatedsteamlinebreakevent,additionaltubes(those with lighest voltage indications) will be removed from service until the ,

primary to secondary leakage limits are satisfied. Also, a reduction of the 1 allowable primary to secondary leak rate, inherent to the interim criteria, to a maximum of 0.31 gpm for all steam generators, will provide margin to the leakage conditions previously evaluated in the V. C. Summer FSAR.

The basis which supports application of the proposed. interim criteria is contained in WCAP-13522. The technical justification for the 1.0 volt bobbin probe signal amplitude interim plugging criteria discussed in this document utilizes both pulled tube and model boiler test specimens. Additionally, the statistical interpretation of the data and the overall methodology are consistent with the industry approach forroulated by the Electric Power ResearchInstitute(EPRI).

Purp?se for Chanag - This proposed license amendment is to preclude unnecessarily plugging tubes due to the occurrence of ODSCC at the tube support plate elevations. The proposed interim tube plugging criteria for Cycle 8 utilizes a conservative bobbin probe voltage-based plugging limit.

This criteria has demonstrated that, in the limiting case, the potential presence of through-wall cracks alone is not reason enough to remove a tube from service. For example, during recent steam generator tube inspections, indications were identified at the tube support plates in each of the 3 steam generators. The eddy current data for ther 'ndications was reevaluated according to " Appendix A" of WCAP-13522, [ . nate Pluaaino Criteria Eddy Current Analysis Guidelines. The data reev% tion indicated that 127 of the-131 indications were found to be below 2 vn1ts in bobbin amplitude, and the remaining 4 indications were between 2 and 3 volts. Therefore, in contrast to the proposed interim plugging criteria, the current Technical Specification steam generator tube plugging limit of 40% tube wall penetration as determined by HDE would cause many of the tubes with crack indications to needlessly be removed from service. The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by personnel involved in tube plugging or repair operations. By reducing nonessential

. . Attachment 2 to Documcnt Control Desk letter T5P 920001 1 Pcge 3 of 20 tube plugging, the proposed amendment would preserve margin to the reactor coolant flow through the steam generator in 0CA analyses. The proposed amendment would also preserve margin to reactor coolant system minimum flow rates and, therefore, maintain an oxcess of that flow required for full power operation. Reduction in the amount of required-tube plugging or repair by sleeving can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.

1

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. . Att: chm:nt 2 to Docum:nt Contral D:sk Lctter TSP 920001 1 Pcge 4 cf 20 Safety Evaluation LICENSING llASIS The plugging criteria involves a correlation between eddy current bobbin probe signal amplitude (voltage) and tube burst characteristics. Leakage during normal operating conditions is not expected to occur upon implementation of the interim plugging criteria. Any potential for primary to secondary leakage during a postulated SLB event is asser, sed relative to conservative leakage values at various bobbin coil voltages. This safety evaluation is completed i'i accordance with 10CFR50.59 criteria and assesses the safety significance of the use of this methodology. RG 1.121. " Bases for Plugging Degraded Steam Generator Tubes", is used as the basis for

<ietermining the adequacy of the new repair limits. This regulatory guide describes a method which is acceptable to the NRC staff for supporting the intentions of General Design Criteria 14, 15, 31, and 32. This method minimizes the probability and consequences of a steam generator tube rupture by determining the safety limiting conditions of degradation of steam-generator tubes, beyond which they must be repaired by sleeving or plugging.

The NRC Regulatory Guide 1.121. " Bases for Plugging Degraded PWR $6 cam Generator Tubes", addresses tubes with through-wall cracking. The Regulatory Guide utilizes safety factors on loads for tube burst and collapse that are consistentwithSectionIIIoftheASMECode.PerparagraphC.3.d(1)ofR.G.

1.121, the applicable analytical and loading criteria in thinned or unthinned tubes with through-wall cracks are;

1. Through-wall cracks in tubes should not propagate and result in tube rupture under accident condition loadings.

2. The maximum permissible crack length of the largest single crack should be such that the associated burst pressure is at-least 3 times the normal operating pressure differential.

3. The leakage rate limit under normal operation set forth in the plant technical specifications should be less than the leakage limit determined for the largest permissible crack.

In establishing the interim tube plugging criteria, the general approach is to verify the existence of acceptable margins to tube burst and excessive steam generator tube leakage during all plant conditions.

The V. C. Summer Technical Specifications have as a basis that for operating leakage up to the specification limit, the dose contribution from the tube leakage will be limited to a small fraction (10%) of the 10CFR100 guidelines in the event of an accident which models secondary steam release to the environment. The current V. C. Summer Technical Specifications limited total steam generator tube leakage to 1.0 gpm (1440 gpd) and 500 gpd through any one steam generator. Per the V. C. Summer FSAR, a postulated locked rotor event results in the largest off- site doses of accidents which model steam release. Using the conservative analyses results from the FSAR, the current thyroid and whole body doses are 9 Rem and 0.6 Rem, respectively, for 1.0 gpm

. . Attachm:nt 2 to Docum:nt Control D:sk letter TSP 920001-1 Pcge 5 of 20 total primary to secondary leakage. The postulated SLB event represents the next most limiting transient relative to potential off-site dose estimates.

The V. C. Summer FSAR evaluation for a main steam line rupture outside of the containment structure indicate whole body and thyroid doses of 0.011 and 1.1  :

Rem, respectively, at the site boundary. The above values represent 0 to 2 l hour dose estimations.

The interim tube plugging criteria also includes the conservative assumption >

that the tube to tube support plate crevices are open (negligible crevice '

depositsorTSPcorrosionexists)andthatthetubesupportplatesare displaced during SLO conditions. The open crevice assumption leads to maximum leak rates compared to packed crevices and also maximizes the potentialforTSPdisplacements(whichcanuncovercracks)underaccident condition loadings. The 0D500 is thus considered to be free span degradation under SLB accident conditions and the principal requirement for tube plugging considerations is to provide margin against tube burst per R.G. 1.121.

Limiting the V. C. Summer Technical Specification operating leakage limits to 0.31gpm(450gpd)totaland0.1gpm(150gpd)persteamgenerator,will maintain the current FSAR analyses results for a locked rotor event (the limiting case for radiological releases) and provide margin to these results.

In support of the ' defense in depth' philosophy for steam generator tube integrity, the 150 gpd limit enhances the leak-before-break characteristics of potentially degraded tube support plate intersections. Per NUREG-0717

" Safety Evaluation Report for the Operation of the Virgil C. Summer Nuclear Station", off-site doses resulting from a main steam line break bound the FSAR results while remaining below the limit of 10% of the 10CFR100 guidelines for primary to secondary leakage of up to 1.0 gpm in the faulted loop.

Similar steam generator tube plugging criteria have been applied for and accepted by the NRC staff on an interim basis for' application to both 7/8" and 3/4" diameter tubing. The V. C. Summer request represents the second time that a license amendment request has been made for a bobbin probe voltage-based plugging criteria for indications at the tube support plate intersections of 3/4" diameter by 0.043 wall steam generator tubing.

However, this request varies slightly from the previous request in that it is necessary for only one cycle, since the V. C. Summer steam generators will be replaced following Cycle 8.

EVALUATION Eddy Current Inspection Capability The proposed license amendment request assesses tube integrity based on standard bobbin coil signal amplitude for dispositioning flaw like indications. Bobbin coil voltage measurement is related to volumetric wall loss and is dependent upon parameters such as crack length, crack opening area,and'integratedcrackintensity(?resenceofligaments).Thus, voltage may be regarded as an " integrated cracc severity index" relative to tube integrity considerations. In the proposeo criteria, the RPC probe is used to verify the mode of degradation occurring at the support plates and can be used as an additional tool for confirming whether higher voltage indications

I o . Attcchm:nt 2 to Docum:nt Control D:sk Lcttcr TSP 920001 1 Pcge 6 of 20 represent actuel tube wall degradation. The standard bobbin probe continues  !

to be the main probe utilized for tube plugging or repair determination.  ;

The probability of detection of indications at tube support plate elevations ,

has been previously uvaluated for both bobbin coil and RPC p' robes. This '

evaluation combined field inspection results with pulled tube examination i results for TSP indications.

The overall pulled tube data base for IGA / SCC indications shows voltage levels as high or higher than obtained for 00500 at comparable depths of indications. Degradation of this type is considered to be readily detectable. Pulled tubes from European plants with more significant levels of ,

IGA exhibited bobbin voltages as high as 11 volts. The burst characteristics cf these tubes were calculated to exceed the Reg Guide 1.121 minimum limit of 3 times the normal operating pressure differential.

Based on the previous evaluation, the probability of detection for stress corrosion cracking macrocrack networks >40% average (actual) depth approaches 100%. However, very short (approximately 0.2") cracks may have a detection threshold of greater than 40% depth. It should be noted that the Technical Specification 40% plugging limit represents an NDE determined depth, dependent on tube wall degradation type..which may represent actual tube degradation in the range of 55% to 60%, based on an assumed uncertainty of 15% to 20% for phase angle based depth calls.

Tube Wall Degradation __ Characterization In general, the degradation morphology occurring at the tube support plate intersections at plants in the U.S. can be described as axial 005C0. Axially oriented macrocracks can occur at one or more azimuthal locations around the circumference of the tube. The macrocracks are comprised of short, nearly collinear microcracks separated by ligaments of material. Typical microcrack length is less than 0.2 inch. The corresponding macrocrack can be as long as the support plate thickness.

Hinor to moderate inter-granular attack (IGA) can occur in addition to the axial ODSCC. To date, 14 tubes have been removed from domestic plants with Hodel D (3/4" outside diameter x 0.043" wall) steam generators and destructively examined. The corrosion morphology of these pulled tubes is consistent with the TSP degradation morphology experienced industry wide. The maximum length of inter-granular corrosion evidenced from these pulled tubes is 0.5 inch, centered about the midpoint of the tube support plate.

Portions of two tubes, R36C67 and R35C70, from V. C. Summer S/G B were removed in October of 1988. The tube sections were destructively examined to ,

characterize the degradation mechanism. The tube support plate intersections were not burst tested.- The results of the tube exam indicated that the degradation mechanism was. axially oriented ODSCC with minor IGA components.

The maximum extent of crack penetration was determined to be 14% through-wall. No wastage or general attack of the tubes was detected. Three influencing factors indicate that the corrosion degradation morphology-occurring at the TSP intersections is consistent with the industry. They '

are: 1) bobbin coil voltages are comparable to other plants with extensive tubedestructiveexamination2)voltagegrowthratesforSummerare

. . Attachment 2 to Docum:nt Control D:sk Lott:r TSP 920001 1 Page 7 of 20 comparable with other plants, and 3) the rotating pancake coil (RPC) inspection history for Summer indicates that the corrosion morphology is in fact outer diameter stress corrosion cracking.

3/4" Model Boiler Burst and Leak Data A set of 47 laboratory induced stress corrosion cracking tube intersections are used in the 3/4" tubing data base in the development of the interim plugging criteria. 1 r. !dition to being taken to true burst, the model boiler test specimens were leak tested at normal operating and steam line break pressure differentials. A summary of the model boiler results are furnished below o Bobbin voltages up to 5.7 volts, with a through-wall crack length of 0.08" did not leak at SLB conditions. Through wall crack lengths up to 0.17" did not leak, o A minimum bobbin coil voltage of 4.24 volts resulted in-leakage at SLB conditions.

o Model boiler test results also indicate that the bobbin voltage corresponding to a burst pressure of 2880 psi (approximately equal totheSLBpressuredifferentialof2650 psi)is10 volts.

o Tubes that burst in a range near the RG 1.121 limit of 3 times normal operating pressure differential (3996 psi for V. C. Summer) varied from 14 to 19 volts. The range of burst pressures for these tubes was approximately 3400 to 4400 psi.

Operatina Plant Leakage Data N11ed tube examination results from other plants indicate that through-wall cracks can potentially occur below 10 volts but that the associated crack lengths are short with no measurable leakage at operating conditions, teakage at operating conditions has not been identified in the field for bobbin coil voltages less than 7.7 volts in a 3/4" tube.

A total of three tubes have been suspected to contribute to steam generator leakage during normal operating conditions where the defect was attributable to ODSCC in the TSP intersection. Two of tnese tubes were discovered at one plant in a operating cycle that experienced leakage of less than 140 gpd.

The two tubes had signal amplitudes of about 10.5 volts and 34 volts (in Europe) and were included with 3 other tubes that had experienced roll transaction cracking, all of which are suspected to have contributed to the leakage. The third tube as discovered at a plant that experienced 63 gpd leakage including other tubes with roll transition cracking. This tube had a signal amplitude of 7.7 volts and an indicated depth of 92% through-wall. No field leakage has.bcen reported for indications below 7.7 volts. Also, no measurable leakage was observed in a Model D Steam Generator tube that indicated, via destructive examination, a through-wall creck of 0.01" long (NDE signal amplitude of 1.90 volts). Therefore, a reasonable judgment for the operational leakage threshold is 7 volts or more.

i .

Attachment 2 to Document Control Desk lettGr TSP 920001 1 Page 8 of 20 Operatina Plant Burst Data A total of 14 pulled tube intersections are included in the overall data base. The pulled tube data base includes tubes from furopean plants, where tube plugging criteria permit much larger voltage indications to remain in service. The pulled tube burst data shows that for indications which represent EOC voltages of 2.0 to 3.0 volts, which is the upper bound of predicted E00 voltages when using a 1.0 volt plugging limit, burst pressures of approximately 5,400 to 7,000 psi can be expected. This is far in excess of the RG 1.121 burst requirement o' 3 times the normal operating pressure (3996 psi). For burst pressures which approached 3 times the norma.1 pressure differential, bobbin voltages ranged from 6.55 to 15.7 volts.

Agcident Condition Leakage Threshold The proposed interim tube plugging criteria addresses the licensing basis limit for SLB leakage of s 1.0 gpm in any one loop, by making a conservative determination of the expected end of cycle SLB leak rate. This determination is based on the voltages of r. rack like indications left in service and the SLB leakage threshold derived from the pulled tube data base. Of the Model Boiler and pulled tube data base for indications below 2 volts (approximately 100 indications), only one case leaked under SLB conditions. This tube was removed from a plant with 3/4" 0.D. by 0.043 inch wall thickness, had a pre-pull bobbin coil voltage of this tube was 1.13 volts and leaked at approximately 0.023 1/hr, or 0.0001 gpm at bounding SLB conditions of 2650 psid. Also, since a large number of lower voltage indications ( < 3 volts) examined to date have not leaked at SLB conditions,the discovery and characteristics exhibited by this tube supports-the probability of leakage correlation discussed later. The leakrate from this indication, when combined with its probability of leakage at SLB conditions, suggests that large_ numbers of_these indications would be necessary_for_their integrated effect to-influence of'-site doses during an SLB. Therefore, a reasonable judgment for the SLB leakage threshold is approximately 1 to 2 volts.

A higher voltage indication of about 2 volts (2.06 volts actual) was found to leak at a rate of 0.17 1/hr (0.00075 This indication represents the threshold of leakageabout (gpm) at SLB 0.001 gpm)conditions.

for SLB conditions that could be expected to contribute to-the off-site dose prediction for a postulated SLB event. assuming many of these indications existed and leaked.

The pulled tube data base demonstrates that indications in the range of 6 volts did not leak at SLB conditions. This supports that 2.0 volt indications have a low probability of any leakage, and when combined with their-potential leakrates, could not produce excessive steam generator -

leakrates. Therefore, a bobbin voltage cased plugging criteria which predicts E0C voltages under 2.0 volts, regardless of the number of 2.0 volt indications, would not be expected to result in primary to secondary leakage of a sufficient magnitude'(> 1.0 gpm) that FSAR calculated off-site dose '

estimations would be exceeded.

q

o a .

Attachment 2 'to Document Control Desk Letter i TSP 920001 1 Pege 9 of 20 lobe Intecrity Assessaena ative to the P.G 1. 1 requirement to ensure cracks do not propagate and uit in tube rupture under r,ccident conditions, the following postulated

.nts were considered: Loss of coolant Accident-(LOCA), Steam Line Break-(SLB),andFeedlineBreak(FO: combination with Safe Shutdown Earthquake (SSE). It is shown that pott.3 C throtgh-wall cracks which may exist as a result of implemen ution of tm i s Jrim tube plugging criteria are not expected to pm W te and resu r. in tubr. rupture under accident condition.

loadings. Thc Je7tial effects of a postulated LOCA + SSE event on continued tubt integrity c7d the ebility of the V. C. Summer steam generators-to perform their intended safety function are addressed below and shown to be acceptable. By th6 conclusion that the combined maximum tube bending stress-at any elevation during a SSE event is less than the tube material yield strength (using lower tolerance limit properties). The potential effects of a postulated SLB + SSE or FLB + SSE event on continued steam generator tube integrity are enveloped by the RG 1.121 criterion requiring a factor of safety of 3 times ncrmal operating pressure differential to tube bun t with the presence of a through-wall crack. Outer diameter stress on the order of the yield strength of the tube material is required before any signif.icant effect on tube burst strength is realized (see WCAP-7832A, Evaluation of Steam Generator Tube. Tube Shert and Divider Plate Under Combined LOCA and_

SSE Conditions). Tube burst capah'111ty following a combined SLB + SSE or 7LB-

+ SSE event upon implementation of the interim plugging criteria is evaluated-below and is found to be acceptable.

Combined LOCA + SSE Loadings - In addressing the combined effects of LOCA +.

SSE c' the steam generator component (as required by GDC 2, and AG 1;121), it has been postulated that local tube collapse may occur in the steam-generators at some plants. The tube support plates may become deformed as a result of radial loads at the wedge supports at the periphery of the plate due to combined LOCA and SSE loadings.

There are two issues associated with local steam generato, tube collapse.

First, the potential collapse of steam generator tubing adjacent to wedge groups reduces the RCS flow area through-the tubes, which may . m te LOCA Peak Clad Temperature (PCT). Second, there is a potential thn to ial through-wall cracks in tubes could progress to through-wall crac!S , or chat existing through-wall cracks might open up, auring tube deformation resulting-in secondary to primary in-leakage,~shich similarly may cause an increase in PCT.

Large break forces were used for the LOCA + SSE evaluation._ However,.a leak-before-break evaluation of the reactor coolant loop piping for V. C. Summer-has been reviewed and approved by the NRC; therefore, the results of this analysis will be very conservative. The results of this analysi' indicate that for the flow distribution baffle elevation, no tubes'would be excluded from IPC iraplementation. However, a limited number of tubes adjacent _to the wedge groups for'the remainder of the plate locations, may deform or collapse and secondary to primary in-leakage may occur. The affected tubes for each TSP location are listed Li WCAP-13522. The IPC cannot be applied to these locations, and they should be plugged or repaired by sleeving if TSP L

-' - Attachment 2 to Document Control H k letter TSP 920001-1 Page 10 of 20 degradation is detected at elevations other than the flow distribution baffle (FDB) in these tubes.

For all other steam generator-tubes (tubes not located in wedge areas), the possibility of secondary to primary leakage in the event of a LOCA + SSE event is not significant based on the following reasons. Any potential secondary to primary leakage is expected to be much less than the' current primary to secondary operating leakage limit. Steam generator tube integrity and operability are enhanced with the reduction of allowable leakage from 500 gpd to 150 gpd per steam generator. Secondary to primary leakage would be less than primary to secondary leakage for the same differential pre =="re since the cracks tend to tighten under secondary to primary differt f pressure. Additionally, the presence of the tube support plate is exp w ed to reduce the amount of in-leakage as the annulus between the TSP hole and the tube functions as a leak limiting orifice.

Combined SLd + SSE Loadings - During this postulated accident, the lateral support provided to the tube by the TSP can induce bending stresses at the TSP intersection, which vary from tension to compression around the tube circumference. Compressive stresses have the potential to reduce the tube burst capability due to crack opening.

Test results show that bending stresses on the order of the tube' material yield strength at operating temperature are required to significantly affect tube burst characteristics. Per RG 1.121, tube burst protection during accident conditions is required to be afforded at a differential pressure of 2650 psi. Also per RG 1.121, the bending stress due to SSE must be coupled with the pressure induced primary membrane stress when evaluating the effects of a combined SLB + SSE event. Based on the results of the seismic analyses, the maximum bending stresses at the top W 'a e been shown to be less than the tube yield strength at temperature, k , "~ed above, bending stresses approaching the material yield are regt it .: u affect burst. Therefore, the postulated effects of a combined SLB + R not adversely affect the burst capability of the V. C. Summer tuning.

Tube Burst Protection - The criterion of RG 1.121 to maintain a factor of 3 times normal operating pressure differential on tube burst is inherently satisfied during normal operating conditions. Based on V. C, Summer eddy current data, the tube support plate elevation ODSCC is situated within the thickness of the tube support plates. Steam generator tube denting (due to TSP corrosion) and cracking potentially initiate and progress at high temperature within the TSP crevice. During normal operation the TSP remains adjacent to the original crevice area. Therefore, any cracking which occurs-at the TSP intersection will be situated within the plates, thus precluding tube burst during normal operating conditions.

An algebraic relationship between bobbin voltage and burst pressures of pulled tube and model boiler specimens was developed using a least squares regression analysis. This relationship plots burst pressure against the log of bobbin probe signal amplitude. Once the burst data was plotted, a one sided, lower 95% prediction interval was calculated. This curyn is further reduced to account for a reduction in the material properties at operating temperature. Additionally, the flow stresses of the tube samples have been normalized to provide further consistency in the data base. The lower 95 %

- . Attachment 2 to Document Control Dask Lett:r TSP 920001 1 Pag 211 of 20 prediction interval is a lower bound curve for a particular data set toat is based on the spread of the data about its mean. The next identified indication is predicted to fall above the lower 95 % prediction interval, based on the variances of current data set. This statistical reduction of the burst data will accurately account for normal data fluctuation. As more data points are included in the data set, the prediction interval will migrate towards the mean, provided the future data is consistent with the current spread. It is judged that the broad range of morphologies observed to date, is representative of operating plant, and with the spread of these burst data, suggests that the lower 95 % prediction interval curve is quite conservative.

Baced on the existing 3/4" data base for free span tubing, and correcting for the influence of V. C. Summer operating temperature on material properties and minimum strength levels, the criteria for tube burst margins per RG 1.121 during both normal and accident condition loadings for V. C. Summer can be W isfied with bobbin coil indications with voltage levels less than 3.7 7ea s, independent of the depth measurement. Alternating crack morphologies (i.e., single or multiple cracks) could correspond to 3.7 volts so that a unique crack length is not defined by the burst pressure to voltage correlation. The lower 95 % prediction interval adjusted to lower tolerance limit material properties bounds the results obtained for the entire spectrum of ODSCC morphologies identified by destructive examination. For EOC voltages below about 8 volts, the closest data point to the lower 95%

• prediction interval LTL curve lies approximately 500 psi above the curve.

Again, conservatism is added by assuming the presence of the tube support plate would not preclude tube burst and the indication will behave as-free span degradation. This methodology for the statistical evaluation of the valtage/5urst data has been reviewed an agreed upon by the EPRI Alternate Plugging Criteria (APC) Task Team. The conservatism of this curve is shown in that the closest data point to the 95 % prediction curve, reduced for material properties, lies approximately 500 psia above the curve.

Development of kepair Limit Based on 3.7 Volt Structural Limit.- A beginning of cycle (B00) repair limit based on the E00 structural limit of 3.7 volts can be established. This repair limit, when combined with-a crack growth allowance and eddy current uncertainty allowance would be expected-to result in E0C voltages at the structural limit. To address the potential for variations in future cycle length, en average, enveloping growth allowance of 45% is applied to establish the repair limit. The average growth allowance is determined from the eddy current results of the previous cycle. These results for V. C. Summer show a 44% average growth (0.29 volt) over the entire range of voltages and about 16% average growth for indications above 0.75 volts at BOC.

The total NDE uncertainty on bobbin probe voltage include such factors as probe wear, analyst variability, and measurement repeatability. The eddy current uncertainty allowance in measuring voltage is conservatively bounded at 20 %. This NDE uncertainty allowance factors in both the contribution from analyst variability and probe wear. During the upcoming outage, eddy-current inspection and analysis guidelines (Appendix A of WCAP-13522) implemented will be consistent with the guidelines developed for other previous alternate tube support plate plugging criteria implementation submittals. The ASME standards used during the upcoming inspection will be

. . . Attcchment 2 to Document Control Desk Lett:r TSP 920001-1 P ge 12 of 20 calibrated against the laboratory reference standard utilized for the APC database. In addition, the probe wear standard discussed in WCAP-13522 will be used. Use of the probe wear standard during the upcoming outage is expected-to reduce the inherent error of voltage measurement, thereby resulting in lower uncertainty and lower total voltages. The overall eddy current uncertainty of 20% represents a conservative adjustment to the calculated 14% value of the root sum square combination of the analyst and-probe wear uncertainties at 90% cumulative probability. When the allowances for growth and NDE uncertainty are applied to a beginning of cycle repair limit of 2.2 volts, the 3.7 volt structural limit is verified. This repair limit of 2.2 volts represents the plugging limit if a full alternate plugging criteria were to be implemented.

In the development of the V. C. Summer interim plugging criteria, a BOC 1.0 volt indication is projected to E0C-8 conditions by adding projected growth and NDE uncertainty. This EOC voltage is then compared to the structural limit of 3.7 volts, which supports the burst criteria outlined in RG 1.121.

Adding the average voltage growth allowance of 45% plus NDE uncertainty of 20% to a 1.0 volt B0C indication, an EOC indication of 1.65 volts is predicted. The predicted burst pressure for a 1.65 volt E0C indication is 4870 psia, and compares well with the RG 1.121, 3AP limit of 3996 psi. The Table below provides other comparisons to the structural limit voltage at 90

%, 95 %, and 99 % cumulative growth, using the Monte Carlo prediction and also the maximum absolute voltage increase from the last cycle. The burst margins are based on a voltage of 3.7 volts for structural integrity at 3AP (3996 psi) and 16.4 volts at true SLB AP of 2335 psi.

Burst Margins (volts)

Plugging Voltage NDE EOC Limit Growth Uncertainty Voltage g3AP. RG SLB 1.121) 1.0 (avg.) 0.45 .20 1.65 2.05 14.75 1.0 (90% cum.) .14 1.85 1.85 14.55 0,71 1.0 (95% cum.) .20 2.10 1.60 14.30 0.9 1.0 Monte Carlo * ---

2.97 0.73 13.47 1.0 (99% cum,) .25** 3.05 0.65 13.35 1.80 1.0 Maximum 2.0 .25** 3.25 (0.45) 6.05

• Monte Carlo results represent the maximum voltage for an approximate 99.5% cumulative distribution for 200 tota's indications. The Monte Carlo prediction includes allowance for growth and uncertainty.

• • An uncertainty of 0.25 volts corresponds to an approximate 99%

cumulative probability for eddy current uncertainty.

l

Attachment 2 to Document Control Desk' Letter  !

TSP 920001-1 Page 13 of 20 For a bounding SLB AP of 2500 psi, per the V. C. Summer FSAR, the voltage corresponding to burst is approximately 13 volts, and large margins would be provided for all possible E0C voltages.

The maximum predicted EOC voltage using the Monte Carlo sampling plan with 200 indications at 1.0 volt BOC in the sample population is 2.97 volts, and -

includes growth and NDE uncertainty allowances. It should be noted that the maximum case (3.25 volts E00) is not derived using any BOC voltage-distribution. For this case, the maximum observed growth from the previous cycle eddy current results is combined with a 1.0 volt BOC indications and a bounding uncertainty. For comparison only, the maximum deterministic _ case of 3.25 volts is compared to the lower 99% prediction interval voltage / burst correlation of 9.3 volts at a AP of 2335 psia. The Monte Carlo prediction compares well with the maximum obtained voltage of 2.81 volts observed from the last eddy current inspection (2.81 volts, 0.81 volts B0C). ,

Definition of Interim Tube Pluacina Criteria Eddy Current Criteria - The V. C. Summer steam generator interim tube plugging criteria for the tube support plate elevation 00 SCC is summarized in Table 1.0. The recommended tube plugging criteria is based upon bobbin coil inspection voltage signal amplitude, which is correlated with tube burst capability. The criteria is developed to preclude free span tube burst under SLB accident condition loadings. Following the guidance provided by the applicable Regulatory Guides, a B0C repair limit of 2.2 volts is established.

This limit is expected to result in E0C bobbin coil voltages bounded by the structural limit of 3.7 volts. For conservatism, an interim plugging limit of 1.0 volt is established. This plugging limit provides an inherent 1.2 volt plugging margin. For indications with bobbin coil signal amplitudes >

1.0 volt but less than 2.2 volts, a RPC inspection of the indication can be performed. If the RPC does not identify ODSCC, the indication can be evaluated by current criteria and possibly remain in service. If an alternate degradation mechanism (wear, thinning, etc.) is detected, the operability of the indication will be assessed using the current 40% by HDE depth based-criteria.

SLB Leakage Considerations - Of the number of tube intersections with voltage signals that satisfy the interim tube plugging criteria, it is conservatively assumed that_some of this number will have the potential for leakage during a postulated steam line break. Leak rate testing of pulled tubes has not demonstrated a potential for significant SLB leakage of a magnitude that.

would contribute to off-site doses from crack > indications with voltages under 2.0 volts. However, all E0C voltages greater than 1.0 volt will be included in the leakage calculation. Uncertainties in voltage signal and growth allowance are accounted for using Monte Carlo techniques and an end of cycle-voltage population is assessed for the potential for leakage during a postulated steam line break event. A deterministic bobbin voltage-leakrate correlation is applied to calculate the potential SLB leakage at E0C-8 for V.

C. Summer. The leak rate data for the model boiler and pulled tube specimens were analyzed in order to establish an algebraic relationship that could be used-to predict both the probability of leakage and leakage itself as a

• - I Attachment 2 to Document Control Desk letter TSP 920001-1 Pcg314 of 20 function of bobbin voltage amplitude._ Data from non-leaking intersections-was not included-in the voltage leakrate correlation.

SLB leakage data is analyzed.using a linear regression analysis. The' slope' of the SLB leak rate correlation'is affected by the selection of the regressorvariable(voltagevs.leakrateorleakrate'vs.. voltage). The final-correlation combines the portions of the two regression curves to obtain a conservative relationship over the entire voltage range. 'Below about B volts, the mean leak rate correlation has a fairly flat slope,'which is representative of the pulled tube Model Boiler data. Above about 8 volts the slope of the mean leakrate correlation line exhibits a steeper slope. This again is representative of the previous discussions of the various leakrate thresholds. The leakrate correlation is exceptionally conservative in the lower voltage range (below 3 volts) when compared to a plot of the raw test data for leakrate and voltage. In. addition, non-linear regression analysis is used to develop a correlation for probability of leakage as function of bobbinvoltagebasedonSLBleakagedetection(orlackofdetection)from44 -

pulled tube intersections and 41 Model Boiler specimens. Based on the E0C- i voltage distribution established by the Monte Carlo analysis, a particular "

probabilityofleakageforeachvoltagebin(i.e.,eachvoltageincrement)is established. The values for the SLB leak rate are then determined by integrating the result of the entire EOC voltage distribution combination with probability of leakage and SLB leak rate.

Applying this methodology to the V. C. Summer projected E00-7 voltage distribution with the probability of leakage and SLB leak rate vs. voltage correlation results in a maximum leak rate of 0.02 gpm for loop B, the most limiting steam generator. The 0.02 gpm leak rate compares favorably with the-accident analyses assumptions of 1.0 gpm in the affected steam generator as identified in Table 15.3 of the V. C. Summer Safety Evaluation Report. By '

assigning a probability of leakage to each particular voltage and conservatively combining the two portions of the leakrate regression curves thus allowing for-leakage from small voltage indications, the leakage-contribution (albeit negligible) from small voltage indicat!:ns below the 2.0 volt threshold are accounted for. The probability of-leakage at 2.15 volts E0C is 0.309 at a one sided lower 95 % confidence band. Combining this probability with the SLB leakrate results in an overall contribution of-0.00093 gpm per 2,15 volt indication. The Monte Carlo analyses indicates a maximum EOC-7 projection of 2.97 volts.

Upon implementation of the interim criteria, the leakrate prediction analysis.

will be performed using the actual E0C-7' voltage distributions to establish the predicted EOC-8 leakrate. The results of the leakrate calculation for E0C-8 will be compared.to the 1.0 gpm limit. As noted in Table 1.0, if the calculated leakrate is found to be greater than 1.0 gpm in-any loop, the largest voltage indications must be removed from service until the calculated leakrate is reduced below 1.0 gpm. Based on the leakrate prediction using;

,E00-6eddycurrentdataprojectedto-EOC-7(0.02gpm),leakratesfarbelow.

1.0 gpm would be expected.

The leakrate calculation is performed using a primary to secondary AP.of 2335 psi. It has been determined for V. C. Summer that the probability of achieving a AP of 2335 psi during the SLB event is 3.3 x 10-6. Upon initiation of the accident, primary pressure drops concurrently with

- - ' Attachm:;nt 2 to Docum:nt Control D:sk lettcr TSP 920001-1.

Page 15 of 20 secondarypressure(ofthefaultedloop). Safety Injection (SI) activation repressurizes the primary system. If the plant operators fail to terminate SI, and allow the primary system pressure to reach the PORV setpoint, primary system pressure will be limited to 2350 psia, resulting in a pressure differential of 2335 psi.

TABLE 1.0 INTERIM STEAM GENERATOR TUBE PLUGGING CRITERIA FOR TUBE SUPPORT PLATE ELEVATION ODSCC Bobbin Signal Voltage Action s 1.0 Limited *

> 1.0 but less than 2.2 RPC**

a 2.2 Plug or Repair Allowable Priamary to Secondary Leakage Limits During Normal Operation- ,

1 Per Steam Generater Total

==== a 150 gpd**

• 450 gpd '

• If it is found that the potential for steam line break leakage at end of cycle conditions for tubes planned to be left-in service exceeds 1.0 gpm in any steam generator, then additional tubes will be plugged to reduce steam line break leakage potential in that steam generator:

to below 1.0 gpm. If additional tubes are to be plugged or repaired by sleeving in order to meet the 1.0 gpm leakage criteria, the largest bobbin coil voltage flaw indications would be plugged or repaired.

• • Plug or repair by sleeving if RPC identifies an ODSCC-indication.

• • • Limiting allowable primary to secondary leakage through degraded tubes helps to maintain leak-before-break, and allows for safe plant shutdown prior to critical crack lengths for tube burst being achieved.

4 >

Attachment 2 to Document Control Desk Letter

- TSP 920001 1-Page 16 of 20 Additional Measures Leakage Potential During RCP Locked Rotor Event - In addition to the postulated SLB event, the impact of implementation of-the IPC upon the radiological consequences of a Reactor Coolant Pump Locked Rotor Event is considered. Per the V. C. Summer FSAR, the Locked Rotor Event results in the largest calculated off-site doses. The large off-site doses for this event are a result of assumed fuel failures during the event, not primary to secondary AP.

Per the V. C. Summer FSAR, the maximum obtainable primary system pressure during the event is 2606 psi. The maximum pressure is achieved at approximately 2 seconds into the event and within 6 seconds the primary system pressure has dropped to below the normal system presstre at the time ,

the shaft seizes, and continJes to drop due to the reactor trip. Secondary system pressure is expected to rise coincidentally with the reduced RCS flowrate. The maximum primary to secondary pressure differential that resultsduringthiseventisapproximately1500to1600 psi (by conservatively assuming the secondary system pressure relief valve is set at 1100 psi). This represents un increase of 168 to 268 psi above' normal operating conditions. Detectable leakage during normal operating conditions for ODSCC at the TSPs has not been identified in domestic or European plants for indications below about 7.7 volts. As the maximum expected end of Cycle 8 voltage for the TSP elevation ODSCC in the V. C. Summer steam generators is 2.97 volts, only a negligible increase'in primary to secondary leakage, if any, would be expected at these pressure differentials upon implementation of the interim plugging criteria. The maximum projected SLB single loop leakage for ECC-7 was predicted to be 0.02 gpm, and this value is expected to bound any leakage during this event. The FSAR analysis assumes that primary to secondary leakage at the current Technical Specification limics (1.0 gpm maximum tor all steam generators) occurs during the locked rotor event.

Since primary to secondary leakage through degraded tube support plate-intersections is not expected to approaco a magnitude of 1.0 gpm during this event upon implementation of the IPC, the current radiological ' consequences of a Locked Rotor event continue to be conselvatively bounding.

Application of a More Restrictive Leak Rate Limit - The current Technical Specification primary to secondary leakage limits are a maximum of 500 gpd (0.35 gpm) in any steam generator and a maximum of 1.0 gpm for all steam generators. 'A primary to secondary leak rate of 150 gpd (0.1 gpm) per steam generator,-0.31 gpm (450 god) total steam generator operating leakage rate-limit'will be implemented to further preclude the potential for excessive leakage during both normal operating conditions and a postulated steam line break event in the V. C. Summer steam generators.

The RG 1.121 acceptance criteria for establishing an operating-leakage limit

[ Item 3 within Section 3.0] requires that plant shutdown be initiated if the leakage associated with the. longest permissible single crack is exceeded and.

thus, provides for leak-before-break. The longest permissible crack is the length that provides a factor of safety of 3 against bursting at normal operating pressure differential. For 3/4" tubing, a voltage amplitude of 3.7 volts for typical ODSCC corresponds to meeting this tube burst recuirement at t

i c Attcchm'ent 2 to Document Control Desk letter TSP 920001-1 Page 17 of 20 the lower 95% confidence level on the burst 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 3 times ,

normal operating differential pressure is about 0.44 inch and for SLB conditions is 0.76 inch. Primary to secondary leakage up to 150 gpd_provides for the detection of a 0.4 inch long through-wall crack, based on mean. leak rate and a 0.6 inch long through-wall crack at the 95 % confidence level.

Thus, for cracking that may be occurring at locations within the TSP elevation, the V. C. Summer 150 gpd limit provides for safe plant shutdown prior to the occurrence crack lengths with a potential for tube burst at SLB conditions. Therefore, the 150 gpd per steam generator leakage limit enhances the leak-before-break protection of degraded TSP intersections above' the current Technical Specification limits.

100% Eddy Current Inspection - Addressing RG 1.83 considerations, upon implementation of the criteria during the Refuel-7 outage, 100% of all hot leg tube intersections down to the lowest cold leg intersection with identified ODSCC will be inspected using the bobbin probe.

An RPC i_nspection will be conducted for all flaw'like bobbin probe _

indications planned to be left in service which exceed a signal amplitude of 1.0 volt but less than 2.2 volts. The RPC results are to be evaluated to characterize the bobbin indications at support plate intersections. If indications other than ODSCC are identified, these indications will be evaluated against a 40% depth criterion for tube plugging / repair. The RPC inspec+: n recommendation is consistent with a threshold value below which SLB leu age is expected to be negligible and other types of degradation (wear, cold leg thinning, etc) are not expected to have a significant effect on steam generator tube integrity.

UNREVIEWED SAFETY QUESTION ASSESSMENT Based on the following justification, operation of the V. C. Summer Nuclear Plant subsequent to implementation of the interim tube plugging criteria.does not involve an unreviewed safety question.

1. Will the probability of an accident previously evaluated in the FSAR be increased?

No. The application of the tube support plate elevation interin tube plugging criteria-is limited to the occurrence of ODSCC within the i

thickness of the tube support plate. Tubes are demonstrated to maintain

! a factor of safety of 3 times normal operating pressure differential for crack indications with signa 1' amplitudes of up to 3.7 volts with no credit taken for potential constraint of the tube support plate under normal operating and postulated accident condition loadings.

O Statistical prediction of E0C voltages upon implementation of the l criteria indicate that margin at EOC conditions will exist between the l'

l l

L i

Attrchment 2 to Document Control D:sk Letter iTSP 920001 1 -

-Page 18 of 20 .

structural. limit and maximum predicted voltage. A single tube rupture event is not postulated.upon impicmentation of the plugging criteria.

2. -Will the consequences of an accident previously evaluated in the FSAR be increased?

No. Although tubes are not expected to burst during accident conditions, it cannot be assured that the cracks will not leak during all plant transients discussed in the V. C. Summer Final Safety Analyses Report (FSAR). Of the accidents that are affected by primary-to-secondary leakage and steam release to the environment, the locked rotor event is most limiting for the V. C. Summer Nuclear Station relative to the potential for'off-site doses. The primary to secondary pressure differential encountered during the locked rotor event is limited.to approximately 1500 to 1600 psi. Tubes left inservice via application of the interim plugging criteria will not result in the potential for an increase in primary to secondary leakage above currently assumed accident analyses during a postulated locked rotor event. The postulated SLB event is the next most limiting transient relative to the potential for off-site doses. The differential pressure encountered during the transient is assumed to result in a potential for increased primary to secondary leakage. Tube intersections-left in service _via application of the interim criteria must be verified to result in projected leakage less than 1.0 gpm in any faulted loop to maintain off-site doses to less than a small fraction of the 10CFR100 guideline. The projected EOC 7 crack distribution is calculated to result in a 0.02 gpm:

primary to secondary 1eakage in the most limiting loop based on the EOC 6 eddy current results. The E0C 7 eddy current results will be evaluated on an equal basis to assure that the leakage prediction does not exceed the licensing basis.

3. May the possibility-of an accident of a different type than any previously evaluated-in the FSAR be created?

No. Neither a single or multiple tube rupture event would be expected in a steam generator in which the interim plugging criteria has been-applied (during all plant conditions). The safety issue associated with-through-wall cracks at the tube support plate elevations is primary to secondary leakage during accident conditions since burst margins are satisfied. The implementation of the interim plugging criteria, coupled with the-identification of a more restrictive normal operating leak rate-limit of 150 gpd, are expected.to result in minimal leakage during a postulated steam line break event.

4. Will'the probability of a malfunction of equipment important to safety-previously evaluated in the FSAR be-increased?

No. The implementation of-the allowable tube repair limits due to the presence of ODSCC in the steam generator tubes at.the support. plate elevations is not_ expected to reduce the overall safety and functional requirements of the steam generator tube bundles._ The steam generator tube bundles will continue to sustain, within recommended margins, the loads during normal operation and the various postulated accident.

conditions without loss of safety function. Conservatively, several

-c ..- Attachment 2 to Docum:nt Control D::sk Letter TSP 920001-1 Page 19 of 20 tube to support plate intersections are to be excluded from the' application of the interim repair criteria for reasons of potential tube collapse or deformation resulting from combined LOCA + SSE-loadings.

5. -Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR he increased?

No. The worst case consequences which could occur during plant operation is primary-to-secondary leakage during normal operating and-plant transient conditions. It has been shown..for the limiting case of a postulated locked rotor event, that the implementation of the interim plugging criteria does not alter the results of-the off-site dose analysis. During a postulated steam line break event, the radiological consequences of projected leakage from the number of tubes in which the plugging criteria has been applied are acceptabl6. For the expected E0C 8 crack distribution in the V. C. Summer steam generators, the consequences do not exceed the currently calculated values, and furthermore do not exceed a small fraction of the 10 CFR 100 guidelines.

6. May the possibility of a malfunction of equipment important'to safety-different than any already evaluated in the FSAR be created?

No. Neither a single or multiple tube rupture event would be expected-in a steam generator in which the repair criteria has been applied (during all plant conditions). The safety issue associated with-through-wall cracks at the tube support plate elevations is primary-to secondary leakage during accident conditions since burst margins are satisfied. The implementation of the repair criteria, coupled with the identification of a more restrictive normal operating leak rate limit of 150 gpd, are. expected to result in minimal leakage during a postulated steam line break event

7. Will the margin of safety as defined in the BASES to any technical.

specification be reduced?

No. As indicated within the above evaluation, following implementation of the interim steam generator tube plugging criteria, the conclusions provided within the FSAR addressing steam generator tube integrity ,

remain valid as current acceptance criteria continue to be met.. Even under the worst case conditions, the occurrence of 00 SCC at the tube support plate elevations is-not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions.-

Typically, the most. limiting effect would be a possible increase':in leakage following a steam-line break event. The potential for excessive leakage during a steam line break event.is minimized by verifying.that the expected distribution of crack indications would result.in less than 1.0 gpm primary to secondary leakage. With this level of leakage.-the radiological consequences are less than a small fraction of the 10 CFR

~

100' guideline. For tubes near tube support plate wedge locations, the possibility of secondary to primary in-leakage in the event of a:LOCA +

SSE event is minimized as the IPC is not applied to locations subject.to potential significant deformation following a LOCA + SSE event.- These; tubes are listed in WCAP-13522. For all other steam generator tubes,-

the possibility of secondary to primary leakage in the event of a LOCA

4

• Attcchm:;nt 2 to Document Control Desk lettsr TSP 920001 1 Pcgo 20 of 20

+SSEeventisalsonotLsignificant;any(ifatall)in-leakageis expected to be less than normal operating leakage (or-less than 150 gpd). In actuality, the amount of secondary to primary-leakage is expected to be less than currently allowed since steam generator tube integrity is being addressed with the reduction in allowed' leakage from ,

500 gpd to 150 gpd per steam generator. Furthermore, secondary to primary leakage would-be less than primary to secondary leakage for the same differential pressure since the cracks tend to tighten under secondary to primary differential pressure.- Additionally, the presence of the tube support plate is expected to reduce the amount of in-leakage. Additionally, the IPC requires 100% inspec+1on of the hot leg tube bundle, plus cold leg intersections down to the lowest cold leg intersection where ODSCC is detected. This requirement far exceeds current Technical Specification requirements, and provides for a greater level of plant safety.

CONCLUSIONS The safety significance of the implementation of a standard bobbin probe signal amplitude interim tube plugging criteria for dispositioning tubes for continued service experiencing stress corrosion cracking at the tube support plate elevations has been evaluated. The implementation of the interiri 1.0 volt bobbin probe signal-amplitude plugging criteria for use at V. C. Summer does not represent an unreviewed safety question in accordance with the criteria outlined in 10CfR50.59.

c c

ATTACHMENT 3 DESCRIPTION OF AMENDMENT REQUEST AND NO SIGNIFICANT HAZARDS DETERMINATION.

e e.

. l

'~

• Att: chm:nt 3 to Document Control Desk Letter TSP 920001-1 Pcge 1 of 12 Description of Amendment Request CurrentLicenseCondition-TechnicalSpecification(TS)3/4.4.5," Steam Generators " states that each steam generator shall be operable in modes 1 through 4. This TS provides further guidance relative to surveillance intervals, steam generator tube sample selection and inspection, and-acceptance and reporting criteria. TS 3/4.4.6.2, " Operational Leakage," ,

provides operational leakage limits for modes 1 through 4. Specifically, the limiting condition for operation (LCO 3.4.6.2.c) allows a 1 GPM total primary to secondary leakage through all steam generators not isolated from the-reactor coolant system and 500 gallons per day through any one steam generator not isolated from the reactor coolant system.

Function of The Affected Technical Specifications - The TS surveillance requirements for inspections of the steam generator tubes relative to TS 3/4.4.5 ensures that the structural integrity of the tubes will be maintained. The p ogram for inservice inspection of the steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

The function of TS 3/4.4.6.2, " Operational Leakage," is to provide a total-steam generator leakage limit of 1 GPM for all steam generators not isolated from the RCS. This ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the' event of either a steam generator tube rupture or steam line break. The 500 gpd leakage limit per steam generator ensures tube integrity is maintained in the event of a main steam line rupture or under LCCA conditions.

Descriotion of Proposed Change - As required by 10CFR50.91 (a)(1), an analysis is provided to demonstrate that the proposed license amendment to implement an interim tube plugging criteria for the tube support plate (TSP).

elevation outside diameter stress corrosion cracking (ODSCC) occurring in the V. C. Summer steam generators involves no significant hazards consideration.

The proposed plugging criteria involves correlations.between eddy current bobbin probe signal amplitude (voltage) and tube burst and leakage characteristics. The plugging criteria is based on testing of laboratory induced ODSCC specimens, extensive examination of pulled tubes from operating steam generators (industry wide), and field experience from leakage due to indications at the tube support plates resulting from axial 00 SCC (identified in European plants only).

The proposed amendment would modify Technical Specifications 3/4.4.5, " Steam Generators," 3/4.4.6, " Reactor Coolant System Leakage," and the associated-bases which provide tube inspection requirements and acceptance criteria to determine the level of degradation for which a tube experiencing ODSCC at the-tube support plate elevations may remain in service in the V. C. Summer steam generators.

For Cycle 8 operation of V. C. Summer, an interim tube support plate elevation plugging criteria is proposed which utilizes a conservative voltage based plugging limit. The interim criteria can be summarized as follows:-

Att chm:nt 3 to Docum:nt Control Desk Letter

' TSP 920001;1 Pcge 2 of 12 Flaw indications with a bobbin coil voltage loss than or equal to 1.0-volt can remain in service without further action. For flaw indications in excess of 1.0 volt but less than 2.2 volts (preliminary repair limit for full implementation of the alternate plugging criteria), the tube can remain in service provided an RPC inspection of the: indication does-not detect a defect. Indications involving other modes of tube degradation will be assessed for operability based on the current 40% depth based, F*, and L* plugging criteria. Crack indications above 2.2 volts will be plugged or repaired by sleeving regardless of RPCresults(ifRPCtested).

To remain consistent with the V. C. Summer licensing basis, potential leakage followingapostulatedsteamlinebreakeventattheendofcycle(E0C) conditions shall be less than 1.0 gpm for the most limiting steam generator.

Off-site doses will consequently remain within a small fraction of the 10CFR100 limits. Bobbin coil flaw indications inspected by RPC and found to have no RPC indications do not need to be included within the' leakage analyses. Bobbin indications less than 1.0 volt, if not inspected by RPC, are to be included in the leakage analyses. If the projected leakage exceeds- -

1.0 gpm during a postulated steam line break event, additional tubes (those with highest voltage indications) will be removed from service until the primary to secondary leakage limits are satisfied. Also, a reduction of the-allowable primary to secondary leak rate, inherent to the interim criteria, to a maximum of-0.31 gpm for all steam generators, will provide margin to the leakage conditions previously evaluated in the V. C. Summer FSAR.

The basis which supports application of the proposed interim criteria is contained in WCAP-13522. The technical justification for the 1.0 volt bobbin probe signal amplitude interim plugging criteria discussed in this document utilizes both pulled tube and model boiler test specimens. Additionally, the statistical interpretation of the data and the overall methodology are consistent with the industry approach formulated by the Electric Power Research Institute (EPRI).

Purpose for Change - This proposed license amendment is to preclude unnecessarily plugging-tubes due to the occurrence of ODSCC'at the tube support plate elevations. The proposed interim tube plugging criteria for Cycle 8 utilizes a conservative bobbin probe voltage-based plugging-limit.

This criteria has demonstrated that,-in the limiting case, the potential presence of through-wall cracks alone is not reason enough to remove a-tube -

from service. For example, during recent steam generator tube inspections, indications were identified at the tube support plates-in each of the 3 steam generators. The eddy current data for these indications was reevaluated according to " Appendix A" of WCAP-13522, Alternate Pluquina Criteria Eddy Current Analysis Guidelines. The data reevaluation indicated that 127 of the 131 indications were found to be below 2 volts in bobbin amplitude, and the-remaining 4 indications were between 2-and'3 volts. Therefore, in contrast to the proposed interim plugging criteria, the current Technical Specification.

steam generator tube plugging limit of 40% tube ~ wall penetration as determined by NDE would cause many of the tubes with crack indications to needlessly be removed from service. The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by personnel involved in tube plugging or repair operations. By reducing nonessential.

r, e -

Attachment 3 to Document Control Desk Letter TSP 920001 1 -

Pcge 3 of 12 tube plugging, the proposed amendment would preserve margin to_the reactor-coolant flow through-the steam generator in LOCA analyses.. The' proposed amendment would also preserve margin to reactor coolant system _ minimum flow rates and, therefore. maintain an excess of that flow required for full power operation. Reduction in the amount of required tube plugging or repatr_by

-sleeving can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.

+

t.n .

Attachment 3 t'o Docurnent Control Desk 1.etttr -

. TSP 920001-1i Pcga 4 of 12 ,

NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS .

I.

Tube Integrity Discussion -

In the development of an interim plugging criteria-for V. C. Summer, R.G.

1.121, " Bases for Plugging Degraded PWR Steam Generator-Tubes" and R.G._1.83,;

" Inservice Inspection of PWR Steam Generator Tubes"'are used:as the bases for-determining that steam generator tube integrity considerations are maintained  :

within acceptable limits. RG 1.83 describes a method. acceptable-to the NRC.

staff for implementing GDC 14, 15, 31, and 32 through-periodic inservice-inspection for the detection of significant tube wall' degradation. RG 1.121 describes a method acceptable to the NRC staff for meeting General' Design Criteria 2, 4,-14, 15, 31, and 32 by reducing the' probability and consequences of steam generator tube rupture through determining the limiting-safe conditions of tube wall degradation beyond which-tubes with unacceptable cracking, as established by inservice inspection,.should be removed from service by plugging. This regulatory guide uses safety' factors on loads for..

tube burst that are consistent with the requirements of Section III of the ASME Code. ,

For the tube support plate elevation degradation occurring in the V. C.

Summer steam generators. tube burst criteria are inherently satisfied during-normal operating conditions by the presence of the tube support plate. The presence of the tube support plate enhances the integrity of-the degraded tubes in that region by precluding tube deformation beyond the diameter af the drilled hole, thus, precluding tube burst. Conservatively, no credit is taken in the development of the plugging criteria for the-presence of the' tube support plate during norma 1 operating and accident conditions. : Based on the existing burst test database for 3/4" tubing, the_ safety requirements for- .

tube burst margins during.both normal and accident condition loadings'can be--

satisfied with end of cycle (EOC) bobbin coil signal amplitudes less than 3.7 volts (WCAP-13522), regardless of the depth of tube wall: penetration-_of-the cracking.

The 3/4" inch data base includes model boiler and pulled tube.informatlun.:

The pulled tube data includes tubes from both U.S.:and European plants.-Due to.the methodology used for setup calibration and frequency mix of the, bobbin signals;'stgnificant differences-in U.S. and European ~ voltage amplitudes have been encountered. A recent study has-concluded that an: appropriate. scaling' factor-to bring European voltages on a consistent basis-with U.S.-voltages is.

approximately 8, and is dependent on the measured voltage. =The; voltage-burst:

curve of~WCAP-13522 includes European voltages scaled appropriately.;:The methodology;used in the European to U.S. voltage normalization is-detailed in:

, WCAP-13522. The methodology has been independently reviewed'and approved by thenuclearindustry.(EPRI). All voltages discussed in the' evaluation and the~ supporting references have-been' appropriately scaled and.should.be considered to be'able to be-referenced to a common standard. ,

Projection of E0C voltages for. tube support plate indications are made using a statistical-sampling plan (Monte Carlo analysis).- The sampling _ plan uses; as its basis the distribution of voltage differences from the previous

, cycle's eddy current results, eddy current uncertainty distribution

c -

Attachment 3 to Document Control Desk Letter TSP 920001-1 Pege 5 of 12 (consistentwithpreviousIPC/APCsubmittals)andthebeginningofcycle-(BOC)indicationvoltagesdeterminedduringthemostrecentexaminationfor-those tubes left in service to which the criteria will be applied. The Monte t

Carlo sampling plan randomly combines the three variables listed above to-obtain a frequency or cumulative probability of E0C voltages. -A maximum EOC voltage can then be defined that properly reflects the number of indications left in service by integrating the tails-of the distribution to one complete indication.

Upon implementation of the plugging criteria, tube leakage considerations ,

must also be addressed. It must be determined that the cracks will not leak excessively during all plant conditions. For the interim tube plugging criteria developed for the V. C. Summer steam generator tubes, no leakage is anticipated during normal operating conditions even with the presence of potentially through-wall cracks. Industry wide the crack morphology of 00 SCC cracks at tube support plate (TSP) intersections, based on destructive examination, is best described as short, tight, axially oriented microcracks:

separated by ligaments of non-degraded material. -The pulled tube data base includes tubes-from-Model D steam generators, and concludes that among Westinghouse steam generators, the morphology of tube degradation at the TSP intersections is independent of steam generator design. No primary to secondary leakage at the TSPs has been detected in U.S. plants. No field leakage during normal operating conditions has been observed at plants ,

similar to V. C. Summer with crack indications having signal amplitudes less than 7.7 volts.

Relative to the expected leakage during accident condition loadings, the limiting event with respect to differential pressure experienced across the SGtubesisapostulatedsteamlinebreak(SLB) event. The lower bound threshold voltage for leakage, at SLB conditions, based on the presence of through-wall degradation seen in the 3/4" pulled tube data base, is approximately 1.13 volts. The tube used to develop this threshold had a-prepull bobbin amplitude of 1.13 volts, a through-wall degradation length of 0.016 inches,andatSLBconditionshadaleakrateof0.023L/hr-(0.0001 gpm) in the laboratory. While this indication represents a lower bound voltage at which SLB leakage is postulated. the data base supports that E0C voltages in the range of 1 to 2 volts would exhibit a low probability of leakage. Bobbin voltages in the database shows that indications:as high as 6 volts did not leak at SLB conditions. All pulled tube bobbin voltages-utilized in the database are referenced according to prepull bobbin-amplitudes. It is judged that the tube removal process can open cracks or cause ligament tearing between microcracks. This is supported by observation that postpull amplitudes are generally 2 to 3 times higher than prepull amplitudes. The lowest voltage indication from the pulled tube data base which had appreciable SLB leakage is a tube with.an indication of 3.54 volts.

The measured leak rate for this tube was 3.3 L/hr (0.0145 gpm). Similarly, the lowest voltage Mo6el Boiler indication was appreciable SLB leakage was 4.24 volts. The measured leak rate of this indication was 0.33 1/hr. Based on pulled tube data,-the SLB leakage of approximately 0.001 gpm, correlates to a voltage of about 2 to 2.5 volts.

' * 'Attachm:nt 3 to Docum:nt Control Desk lett r

- TSP 920001-1 Page 6 of 12.

Additional Consideratior.s The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by personnel involved in tube plugging or repair operations. By reducing nonessential tube plugging, the proposed amendment would preserve margin to the reactor coolant flow through the steam generator in LOCA analyses. The proposed amendment would also preserve margin to-reactor coolant system minimum flow rates and, therefore, maintain an excess of that flow required for full power operation. Reduction in the amount of required tube plugging or repair by sleeving can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.

10 CFIt 50.02 ANALYSIS Inaccordancewiththethreefactortestof10CFR50.92(c), implementation of the proposed license amendment is analyzed using the following standards and found not to: 1)involveasignificantincreaseintheprobabilityor consequences of an accident previously evaluated; or 2) create the possiLility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in margin of safety.

Conformance of the proposed amendment to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:

1) Operation of V. C. Summer in accordance with the proposed license amendment does not involve a lignificant 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 show burst pressures in excess of 5475 psi for indications of outer diameter stress corrosion cracking with voltagemeasurementsashighas11 volts (WCAP-13522).Bursttesting performed on pulled tubes from similar plants with 3/4-inch tubing shows a measured burst pressure of approximately 5400 psi at room temperature for an indication at 3.5 volts. Correcting for.the effects of temperature on raaterial prope.rties and minimum strength levels (as the burst testing was-done at room temperature), tube burst capability significantly exceeds the RG 1.121 criterion requiring the maintenance of a margin of 3 times normal operatirg pressure differential (3AP) on tube burst. The 3AP for the V. C.

Summer steam generators corresponds to 3996 psi. Bobbin voltages of tubes which burst near the 3AP limit on pressure differential ranged from 8.55 to.

15.7 volts. Based on the existing data base, this criterion is satisfied with 3/4" diameter tubing ~with bobbin coil indications with signal amplitudes less than 3.7 volts, regardless~of the indicated depth measurement. This-structural limit is based on a least squares fit of the burst data'at a lower 95 % prediction interval, further reduced for operating temperature. Bursc tube crack morphologies include tubes with single or only a few-large axial cracks (more representative of Model-Boilers), multiple axial crack-networks which form larger macrocracks, and macrocrack networks which are formed predominantly of axial macrocracks with smaller, shallower circumferential-cracks joining the axial macrocracks. This last morphology has been termed

' -- Att chrn;nt 3 to Docurnent Control D:sk Lctt:r TSP 920001-1.

P ge 7 of 12 cellular corrosion. The burst data reduction conservatively bounds all data points from. destructive examinations performed to date. No crack morphology detected has resulted in burst pressures for a particular voltage which is not representative of the entire data spread and not bounded by the voltage / burst correlation. Therefore, it is judged that the current voltage / burst correlation is representative of any observed or expected (by metallography) tl " degradation morphology affecting the TSP crevice area.

When considering calculated growth rates for ODSCC within the V. C.

Summer steam generators, a 1.0 volt plugging is shown to adequately protect the structural limit. Considering an average voltage increase from the previous cycles eddy current results of 0.29 volts, and adding a 15% NDE uncertainty of 0.15 volts (15% of 1 volt) to the interim plugging criteria of 1.0 volts results in an EOC voltage of 1.44 volts. This end of cycle voltage compares favorably with the structural limit of 3.7 volts. When 90%

cumulative probability values for growth (0.70 volts) and 14% NDE uncertainty values (0.14 volt) are applied to the 1.0 volt plugging limit, the projected E0C voltage of 1.84 volts is obtained. From the voltage / burst correlation in WCAP-13522, the predicted burst pressure of a 1.84 volt indication is 4740 psi,19% greater than the 3AP limit established by RG 1.121 of 3996 psi. The growth rate used to determine the projected E0C voltage is based on the review of growth rates for 87 TSP intersections. This includes all reported indications for which detectable eddy current signals were observed from both the 1990 and 1991 inspections.

Additionally, representing the bounding case._the maximum voltage increase from the previous cycle of 2.0 volts is applied to the 1.0 volt plugging limit, resulting in an E0C voltage of 3.0 volts, which is still below the E0C structural limit. Adding NDE uncertainty at 99% cumulative probability to the maximum growth results in a bounding E0C voltage of 3.25 volts, which is 0.45 volts below the RG 1.121 structural limit of 3.7 volts. It must further be noted that the RG 1.121 structural limit applies a factor of safety of 3 to the normal operating pressure differential._ A more realistic comparison would be to compare this bounding E0C voltage of 3.25 volts to.the voltage which would indicate a burst potential at SLB. Using the lower _99 %

prediction interval burst correlation results in a burst pressure voltage for SLB conditions of approximately 9.3 volts.

The B0C voltage limit of 2.2 volts is used to protect against E0C voltages exceeding the 3.7 volt, RG 1.121 structural limit. The 2.2 volt value represents an upper bound to assess the acceptability of bobbin coil indications, regardless of RPC verification. It is conservatively assumed that B0C voltage levels of this magnitude and greater could result in an indication reaching the structural limit at E0C conditions. The 2.2 volt limit is derived by starting with the 3.7 volt structural limit and reducing it by the NDE uncertainty at 90% cumulative probability (20% uncertainty, .45 volts) and the average voltage growth (45% growth, 1.0 volt). This methodology is considered conservative since the 90% cumulative uncertainty.

value is found to be 14%, but conservatively assumed to be 20% for the analysis. Average growth values for all indications was 44%, while average growth from BOC indications above 0.75 volt was found to be 16%. The 1.0 volt growth allowance (45%) corresponds to 96% cumulative voltage growth. Use

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Att: chm:nt 3 to Docum:nt Control Desk letter TSP 920001 1 P:ge 8 of:12 of a 1.0 volt interim plugging limit implies an inharent 1.2-volt margin to a plugging-limit developed using RG 1.121 methodology.

Only three indications of ODSCC have been reported to have operating leakage

- all three have been in European plants. No field. leakage has been reported at other plants from tubes with indications with a voltage level of under 7.7 volts (from 3/4" tubing). Relative to the expected leakage during accident condition loadirgs, the accident analyses affected.by primary-to-secondary leakage and steam release to the environment as described in the V. C. Nmmer FSAR will not be affected upon implementation of the IPC.

A deterministic bobbin voltage-leakage potential correlation is applied to the Monte Carlo voltage predictions to calculate the potential SLB leakage at the EOC-8 at V. C. Summer. The leak rate data for the model boiler and pulled tube specimens were analyzed in order to establish an algebraic relationship that could be used to predict the probability of leakage as a function of bobbin voltage amplitude. A correlation between bobbin voltage and SLB leakrate has been established. The slope of the SLB leak rate correlation is affected by the selection of the regressor variable (voltage vs.leakrateorleakratevs. voltage). The final correlation combines the portions of the two regression curves to obtain a conservative relationship over the entire voltage range. Based on the protected E0C voltage distribution established by the Monte Carlo analysis, a-particular probabilityofleakageforeachvoltagebin(i.e..eachvoltageincrement, usually kept to 0.1 volts for simplicity) is established. The values for SLB leak rate for each voltage bin are applied to the leakage probability for each bin. The total SLB leak rate is then deter.,ined by integrating the result of the entire EOC voltage distribution.- Applying-this methodology to- <

the V. C. Summer projected E0C voltage distribution from the previous cycle results in a maximum leak rate of 0.02 gpm for loop B,'the most limiting steam generator. The 0.02 gpm SLB leak rate compares favorably with the accident analyses assumptions of 1.0 gpm identified in Table 15.4-23 of'the V. C. Summer FSAR. When the probability of leakage and leaktate at 95 %

confidence limits are applied to a 2.0 volt indication, it is determined that overall effect of each 2.15 volt E0C indication would be to contribute approximately 0.0009 gpm per 2.15 volt indication during an SLB with a AP of 2335 psia. The Monte Carlo analyses indicate a maxium E0C 7 projection of 2.97 volts (99.5% cumulative probability). This value is typical of the-maximum voltages observed following application of the existing depth based criteria.

Upon application of the interim plugging criteria, only a negligible increase in leakage (if any increase at all is determined) above normal operating leakage would be expected during plant transients.

Therefore, as-implementation of the proposed 1.0 volt interim plugging criteria during Cycle 8.does not adversely affect steam generator tube integrity and results in acceptable dose consequences, the proposed amendment does not result in any increase in the probability or consequences of an accident previously evaluated within the V. C. Summer FSAR.

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. x I LAttachment 3 to Document Control' Desk Letter -

c TSP 920001 1 Page 9 of 12:

2)- The proposed license amendment does not create the possib111tyLof:a new or different kind _ of accident from any accident previously:

evaluated.

Implementation of the proposed interim tube support plate elevation steam generator tube plugging criteria does-not introduce any signifl1 cant changes to the plant design basis. Neither a' single or multiple tube _ rupture-event would be expected in a steam generator in which the plugging-criteria has been applied (during-all plant conditions). And use of the criteria does not-

~

provide a mechanism which could result in an accident that has not prev _1ously been evaluated.

Upon application of the interim plugging criteria, no excessive primary-to-secondary leakage is anticipated during all plant conditions due to degradation at the tube. support plate elevations in the V. C. Summer steam generators. However, the RG 1.121 criterion of providing protection against the leakage from the maximum permissible single-crack length which may be achieved during' Cycle 8 operation must'be met.- The pr_imary-to-secondary leckage limits proposed to be implemented with the interim plugging criteria is conservative with respect to the RG 1.121 criterion.

Concurrent with the implementation of the interim plugging criteria, SCE&G will implement a maximum operational leakage limit.of.'450 gpd for-all steam generators, and a maximum of 150 gpd for any one steam generator. The leakage limits will help preclude the potential for-excessive leakage during -

all plant conditions. The 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 crack length that provides a factor of safety of 3 against bursting at normal-operating pressure differential. A voltage. amplitude of 3.7 volts for typical' ODSCC corresponds to this tube burst requirement'at_a lower 95% prediction interval on the burst correlation. Alternate crack morphologies can correspond to 3.7 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:3. times normal operating pressure differential and SLB conditions are 0.44. inch and 0.76 inch, respectively. Nominal leakage for these crack lengths would range from about 0.2 gpm to' 2.3 gpm, 'respectively, while lower 95 % confidence level leak rates would range from about 0.04 gpm to 0.33 gpm, 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 the lower 95% confidence' level leak rates. Thus, the 150 gpd limit provides for plant shutdown prior-to reaching critical crack lengths for SLB conditions at leak rates less than a lower 95 % confidence level and'for three' times normal operating pressure-differential at -less than nominal leak rates.

Application of the 1.0 volt interim steam generator tube plugging criteria at V. C. Summer will help preclude tube burst during all plant conditions during Cycle 8 operation. Tube burst margins are expected to meet or exceed RG c

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Attachment 3 to Document Control D:sk Letter TSP 920001 1 Pega 10 of 12 1.121 acceptance criteria. The limiting consequence of-the. application of the interim plugging criteria is a potential for primary to secondary leakage' below the currently allowable value of 1.0-gpm. Based on_the previous cycle eddy current results projected to E0C conditions, a maximum SLB leakage of 0.02 gpm is predicted. -This amount.of leakage would result in-off-site radiological consequences well within a small fraction'of 10 CFR-100 limits.

Unacceptable leakage is not anticipated at normal operating or reactor coolant pump locked rotel conditions. Therefore, as the-existing tube integrity criteria'and accident analyses assumptions and results continue to be met, the proposed license amendment does not create the possibility of a new or different kind of accident from any previously evaluated. During the Refueling Cycle 7 outage (projected for March 1993) the eddy current results will be evaluated and a specific leak rate calculation for E0C-8 will be generated. Based on the prior eddy current inspection results, the leakrate is expected to be much less than 1.0 gpm.

3) The proposed license amendment does not involve a r.ignificant reduction in margin of safety. ,

The use of the voltage based bobbin probe interim plugging criteria at V. C.

Summer has been demonstrated to maintain steam generator tube integrity commensurate with the criteria of Regulatory Guide 1.121. RG 1.121 describes a method acceptable to the NRC staff for meeting GDCs 14,.15, 31, and 32 by reducing the probability or the consequences of steam generator tube rupture.

This is accomplished by determining the limiting. conditions of degradation of steam generator tubing, as er,tablished by inservice inspection,.for which tubes with unacceptable cracking should be removed from. service. Upon implementation of the criteria, even under the worst case conditions.: the occurrence of ODSCC at the tube support plate elevations is-_not expected to lead to a steam generator-tube rupture event during normal or faulted. plant conditions. The end of cycle distribution of crack indications at_the tube support plate elevations is calculated to result in_ minimal primary to secondary leakage during all plant conditions and radiological _ consequences are not adversely impacted.

It has been determined that the combined effects of LOCA + SSE on the steam generatorcomponent(asrequiredbyGDC2),-mayresultinlocalizedtube deformation in the area of the wedge regions, at the upper support plates.

Analyses results show that for the V. C.. Summer steam generators several ~

tubes-near wedge locations could be affected in this manner. .These tubes.

have been precluded from application of interim plugging criteria (Reference 3). For all other steam generator tubes, the possibility of secondary to -

primary leakage in the event of a LOCA + SSE event is not significant. In actuality. the expected amount of secondary to primary leakage in the event of a LOCA + SSE is expected to be much less_.than the maximum primary to secondary leakage associated Wh the application of this criteria, i.e.,150 gpd per steam generator . Secondary to primary in-leakage would be less than primary to secondary-leakage for the same pressure differential since the cracks would tend to close under a-secondary to primary pressure differential. Additionally, since.no. TSP deflection would be postulated during a LOCA, the presence of the tube support plate is expected to reduce the amount of'in-leakage. Any estimation of.in-leakage is further reduced based on the expectation that E0C bobbin voltages are predicted (by Monte

' ' : 4 -

Attcchment 3 to Document Control Desk Letter

-TSP 920001-1 Page 11 of 12 Carlo) to be a maximum of 2.97 volts, compared to the threshold limit for normal operation leakage of 7.7 volts.

With regard to limits on tube structural integrity as defined by RG 1.121, the proposed V. C. Summer single cycle 1.0 volt plugging limit provides for additional margin against tube burst at EOC conditions compared to the currently used 40% depth based criteria. Previous studies have indicated that the through-wall growth rate for ODSCC at the TSP's to be about 10 to 15% per cycle. When these values are combined with eddy current uncertainty (assumed to be 15% for depth based calls), an indication just below the plugging limit of 40% can result in EOC reported depths of up to about '0%

through-wall. Depth based criteria modeling would permit crack lengths up to-the TSP thickness for which 70% depth provides 3AP capability. This is approximately equal to the structural limit determined by_the criteria of RG 1.121. Using the 1.0 volt criteria, a maximum _EOC voltage of 2.97 volts is predicted using Monte Carlo. This methodology predicts E0C voltages at a-cumulative probability of approximately 99.5 % for 200 indications in the population. The E0C voltage which corresponds with a burst pressure equal to three times the normal operating pressure differential is approximately 3.7 volts, therefore, a 20% margin (0.73 volts) between maximum projected-E0C voltages and the structural limit voltage is afforded using the voltage based criteria, where little or no margin could be expected for a similar candition-tube using the current 40% depth based criteria. Also, comparing the maximum projected EOC voltage with the maximum bobbin-voltage obtained during the last inspection indicates that E0C voltages are approximately equal-for both criteria. The maximum voltage from the last inspection was 2.81 volts. with a growth of 2.01 volts. The maximum E0C voltage predicted by Monte Carlo of 2.97 volts assumed an initial BOC voltage of 1.0,-indicating total growth and uncertainty of 1.97 volts.

Addressing RG 1.83 considerations, implementation of the bobbin probe voltage based interim tube plugging criteria of 1.0 volt is supplemented by: use of a probe wear standard, enhanced. eddy current inspection guidelines to provide consistency in voltage normalization, a 100% eddy current inspection sample size at the tube support plate elevations, and rotating pancake coil inspection requirements for the larger indications left-inservice. Whether a depth based or voltage based criteria is used, the peak-to-peak voltage is-independent of the phase angle calibration for depth and will-inherently _

provide for a more accurate methodology for assessing degraded tube operability. Pulled tube experience, for some crack morphologies has shown that as crack depth and length increase, bobbin voltage increases also.

As noted previously, implementation of the tube support plate elevation plugging criteria will decrease the number of tubes which must be repaired by sleeving or taken out of service by plugging. The installation of steam generator tube plugs reduces the RCS flow margin. Thus, implementation of the interim 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 with-respect to plant safety as defined in the Final Safety Analysis Report or any BASES of the plant Technical Specifications.

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Att chment 3 to Docum:nt Control D:sk lett:r TSP 920001 1 Pcga 12 of 12 CONCLUSION Based on the preceding analysis, it'is concluded that usins the bobbin, voltage-based, TSP interim plugging criteria for dispositioning tubes in the

- V. C. Sunrner steam generators is acceptable and the proposed license amendment does not involve a Significant 11azards Consideration finding as defined'in 10 CFR 50.92, i

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