ML20078P840

Proposed Tech Specs Re Permanent voltage-based Repair Criteria for Outside Diameter Stress Corrosion Cracking at SG Tube Support Plates for Both Units at Plant

ML20078P840
Person / Time
Site:	Farley
Issue date:	02/14/1995
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20078P838	List: ML20078P833 ML20078P840
References
NUDOCS 9502170249
Download: ML20078P840 (26)

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i Revised Technical Specification Pages

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Changes marked with Strikethroughc and Bold, Italics REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2 Tubes in those areas where experience has indicated potential problems.

3.

At least 3% of the total number of sleeved tubes in all three steam generators 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.

4.

A tube inspection (pursuant to Specification 4.4.6.4.a.8) shall be performed on each selected tube.

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

5.

Tube support plate indications left in service as a result of application of the tube appport plate plugging criteria shall be inspected by bobbin coil probe during the following refueling outages.

c.

The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection nay 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.

d.

Inp1=ntation of the steam generator tube / tube appport plate plugging criteria requires 100 percent bobbin coll inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube appport plate with known outside diameter stress corrosion cracking (ODSCC) indications.

The detemina tion of tube appport plate intersections having ODSCC indications shall be based on the perfomance of at least a 20 percent randaan sampling of tubes inspected over their full length.

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-7 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.

Changes marked with Strikethroughc and Bold, Italics l

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

4.4.6.2.2 Steam Generator F* Tube Inspection - In addition to the minimum sample size as determined by Specification 4.4.6.2.1, all F*

tubes will be inspected within the tubesheet region.

The results of this inspection will not be a cause for additional inspections per Table 4.4-2.

I FARLEY-UNIT 2 3/4 4-10 AMENDMENT No. G3,64 l

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Changes marked with Strikethrough and Bold, Italics I

REACTOR COOLANT SYSTEM i

i SURVEILLANCE REQUIREMENTS (Continued) l 4.4.6.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a.

The first inservice inspection 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-1 category or if two consecutive inspections demonstrate that previously observed degradation 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 Specificatio.,

4.4.6.3.a; 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 tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.7,2.

2.

A seismic occurrence 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.

EARLEY-UNIT 2 3/4 4-11 l

Changes marked with Strikethrougho and Bold, Italics REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.4 Acceptance criteria a.

As used in this Specification:

1.

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

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

1 2.

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

3.

Degraded Tube means a tube, including the sleeve if the tube has been repaired, that contains imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

4.

% Degradation means the percentage of the tube or sleeve 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 or sleeve containing a defect is defective.

6.

Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be repaired (i.e.,

sleeved) or removed from service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. This definition does not apply to the area of the tubesheet region below the F* distance in the F*

tubes.

For a tube that has been sleeved with a mechanical joint sleeve, through wall penetration of greater than or equal to 31% of sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by plugging.

For a tube that has been sleeved with a welded joint sleeve, through wall penetration greater than or equal to 37% of sleeve nominal wall thickness in the sleeve between the weld i

joints requires the tube to be removed from service by plugging. This definition does not apply to tube sppport plate intersections for which the voltage-based plugytag criteria are being app 1 Led.

Jtefer to

4. 4. 6. 4.a.14 for the plugging 1Latt applicable to these intersections. At-teb c uppert-p4ete-int e r e c ctioner thc :cpc4+-44fnH-for thc Tenth - Opc rc ting Cyc1 c -- ie-based on-+nalneefting ct~ea ;en:rct r tubc c:rviceebi-14tv-es deeetibed bcicu:

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FARLEY-UNIT 2 3/4 4-12 AMENDMENT NO. 94 i

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Changes marked with Strikethroughc and Bold, Italics REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c.

Degr:d: tion :ttributed t Out:ide di ::ter :treme eecessi:n :::ching xithir th: 5:und ef the tub:

pp;rt pict: with : bobbir v;1teg: greater then 1.0 v;1t will b; : pcired :: plugg:d oncept ce noted ir i t.E.t.:.E.d beleu.

d.

Indicati:n: ef potent-101 degr:d: tion attributed I

t cut:id; di meter :te :: correcica cracking within the 5:und; cf the tube eupport plate with a b;bbir volteg: geeater than 1.0 volt but les:

ther :: equel t: 3.5 colt: m:y ::=:ir ir :: vie if : ::tsting p:n :ke ::11 preb: (PPC) insp : tion d :: nt detect degrad:tien

!ndic ti:n: Of cut:id; di =:t:r tr=== ::::::i = ::: hing degradetion with a b;bbir v ltag: g ::ter then 3.5 volt: uill be plugged er repaired.

7.

Unserviceable describes the condition of a tube er sleeve if it leaks or contains a defect large enough to affect its structural integrity in the event of en Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.6.3.c, above.

i 8.

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 l

cold leg.

For a tube that has been repaired by

~

sleeving, the tube inspection should include the sleeved portion of the tube.

9.

Tube Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-lll78, Rev. 1, or laser welded sleeving as described by Westinghouse report WCAP-12672, which is used to maintain a tube in service or return a tube to service. This includes the removal i

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

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FARLEY-UNIT 2 3/4 4-12a AMENDMENT NO. 94

Changes marked with Strikethroughc and Bold, Italics REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 10.

Preservice Inspection 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, 11.

F* Distance is the distance of the expanded portion of a tube which provides a sufficient length of undegraded tube expansion to resist pullout of the tube from the tubesheet. The F* distance is equal to 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation.

12.

F* Tube is a tube:

a) with degradation equal to or greater than 40% below the F* distance, and b) which has no indication of imperfections greater than or equal to 20% of nominal wall thickness within the F* distance, and c) that remains inservice.

13.

Tube Expansion is that portion of a tube which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the tube and the hole in the tubesheet.

14.

Tube Support Plate Plugging Limit is used fc,r the disposition of a steam generator tube for continued service that is experiencing outside diameter stress corrosion cracking confined within the thickness of the tube support plates.

These criteria are applicable for the Eleventh Operating Cycle only. At tube support plate intersections, the repair limit is based on unintaining steam generator tube serviceability as described below:

a.

Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbia voltage less than or equal to 2.0 volts will be allowed to r===4n in service.

b.

Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than

2. 0 volts stil be repaired or plugged except as noted in 4.4.6.4.a.14.c below.

Changes marked with Strikethroughc and EcId, Italico c.

Indications of potential degradation attributed to outside dianneter stress corrosion cracking within the bounds of the tube stqpport plate with a bobbin voltage greater than 2.0 volts but less than or equal to 5.6 volts anay rma=4n in service if a rotating pancake coil inspection does not detect degradation.

Indications of outside diaameter stress corrosion cracking degradation with a bobbin voltage greater than 5.6 volts will be plugged or repaired, b.

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

4.4.6.5 Reports Following each inservice inspection of steam generator tubes, a.

the number of tubes plugged, repaired or designated F* in each steam genere*or shall be reported to the Commission within 15 days or une completion of the inspection, plugging or repair effoit.

b.

The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in a Special Resort 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 and sleeves inspected.

2.

Location and percent of wall-thickness penetration for i

each indication of an imperfection.

3.

Identification of tubes plugged or repaired.

t FARLEY-UNIT 2 3/4 4-13 AMENDMENT NO. 46 64 7

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Changes marked with Strikethroughs and Bold, Italica REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) t c.

Results of steam generator tube inspections which fall into Category C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

d.

For Laplementation of the voltage-based repair criteria to tube stqpport plate intersections, notify the staff prior to returntag the steam generator to service (Mode 4) should any of the following conditions arise:

1.

If estimated leakage based on the actual end-of-cycle voltage distribution would have exceeded the leak lintt (for the postulated==4 n steam 14n= break utilising 1Loensing basis assumptions) during the prestous operating cycle.

2.

If circumferential crack-like indications are detected at the tube appport plate int:ersections.

3.

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

4.

If the calculated conditional burst probability exceeds 1 x 10'#, notify the 3GRC and provide an assessment of the safety significance of the occurrence.

i FARLEY-UNIT 2 3/4 4-13a AMENDMENT NO. 46,63 i

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Changes marked with Strikethroughc and Bold, Italics REACTOR COOLANT SYSTF){

OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

Fer--the Tenth Operating Cycle only, p Primary-to-secondary l

leakage through all steam generators shall be limited to 450 gallons per day and 150 gallons per day through any one steam generator.

Fer--eub::quent cycles, 1 CPM tot +1--prir.c ry te-secondsey leakage-through-all eteem-genereter: and 500 gallen per day through--any onc stcar generater, d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and i

e.

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

f.

The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve shall be as specified in Table 3.4-1 at a pressure of 2235 i 20 psig.

i APPLICABILITY:

MODES 1, 2, 3 and 4 ACTION:

J a.

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

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, 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 l

leakage greater than the limit specified in Table 3.4-1, isolate the high pressure portion of the affected system from i

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 />.

FARLEY-UNIT 2 3/4 4-17 AMENDMENT NO. 94

Changes marked with Strikethroughc and Bold, Itslics REACTOR COOLANT SYSTEM 3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.9 The specific activity of the primary coolant shall be limited to:

E Lcce-than er equ:1 t: 0.25 micrcCuede per grc DOGE EQUWALENT I-131 for the Tenth Operating Cycic cnly:

bra.

Less than or equal to tre 0.5 microcurie per gram DOSE EQUIVALENT I-131 for cub:cqucnt cycles; erb.

Less than or equal to 100/E microcurie per gram.

APPLICABILITY:

MODES 1, 2, 3,

4 and 5 ACTION:

MODES 1, 2 and 3*:

a.

For the Tenth Oper-eting Cyclc Only,.ith the spccific eebivity of the Freimary ccclant gecatcr than 0.25 micr Cuele pee-eye = DOSE EOUIV?,LE"T I-131 for merc then

'S heurs during en continucue ti=c interval er encccding the limit line ehcu-en Figure 3.'-1, he in et 1ccet "OT ST?."BBY uith Tg 4eee-then 500'T.: Rhin C hcurs.

bra.

For cubecquent cyclce w with the specific activity of the r

primary coolant greater than 4ve 0.5 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous i

time interval or exceeding the limit shown on Figure 3.4-1, be in at least HOT STANDBY with T less than 500'F within 6 avg hours.

l erb.

With the specific activity of the primary coolant greater l

than 100/E ndcroCurie per gram, be in at least HOT STANDBY with Tavg less than 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

• With T greater than or equal to 500'F.

avg 1

i FARLEY-UNIT 2 3/4 4-23 AMENDMENT NO. 94

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Changes marked with Strikethroughs and Bold, It.clics REACTOR COOLANT SYSTEM ACTION:

(Continued)

MODES 1, 2, 3,

4 and 5:

For the Tenth-Operating Cycle only, with th Op :ifi tivity of the prim:ry 0 1:nt gre ter than 0.25 mier:Curi per gram DOGE-EQUIVALENT I-131 :: greater then 100/E micr:Curic per gr:m, perfe:m th ::=pling :nd sn: lysis requirement: Of item 4: ci Table until th Opecifie ::tivity of the primary 00:1:nt is restored to uithir it: limits.

bra.

For subs :;uent cycles, w With the specific activity of the primary coolant greater than -1,4 0.5 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microCuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

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FARLEY-UNIT 2 3/4 4-24 AMENDMENT NO. 94 l

Changes marked with Strikethroughc and Bold, Italics TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSli M 3 GRAM M>

h TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE Q

AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED I

1.

Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3,

4 2.

Isotopic Analysis for DOSE 1 per 14 days 1

N EQUIVALENT I-131 Concentration 3.

Radiochemical for [

1 per 6 months

• 1 Determination 4.

Isotopic Analysis for Iodine a)

Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1#, 2#, 3#, 4#, 5#

Including I-131, 1-133, and 1-135 whenever the specific activity exceeds 4r0 0.5 l

pCi/ gram DOSE EQUIVALENT 1-131 or 100/5 pC1/ gram, wg and f

b)

One sample between 2 and M

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a 1,

2, 3

THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

1. Until the specific activity cf the primary coolant system is restored within its limits.

+ Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

Changes marked with Strikethroughc and Bold, Italico Page Saver l

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FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POVER with the Primary Coolan+, Specific Activity > 1.0 pCi/ gram Dore Equivalent I-131 FARLEY-UNIT 2 3/4 4-26 AMENDMENT NO.

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Changes marked with Strikethroughc and Bold, ItcIlcs REACTOR COOLANT SYSTEM BASES 3/4.4.6 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 maintained. 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 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 manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion : racking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the i

primarycoolantsystemandthesecondarycoolantsystem(primary-to-secondary l leakage = 600 150 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plent: h ce d:=en:tzsted that-peimary t: :::cadary Ic ksg cf 500 gelicas per day per :te = genereter Opezat1onal leakage of this magnitude can be readily be detected by existing j

Tarity Unit 2 radiation monitors of-etes: generat+r--blcudeu.

Leakage in excess of this limit will require plant shutdown and an unscheduled I

inspection, during which the leaking tubes will be located and plugged or repaired.

For--the-Tenth-Opereting Cyel en1y, t The-repa4-t--14mit fc: t-ub t;.1th flew i nd lea t4ene-een te4 *ed-wi thin-the-bounds-of-e-t-ube-euppe rt-plet+-hee-been provided-to-thWG-in-southern Neelear-operet4ng Cc=peny letter dated July Mr4093r The repair limit for ODSCC at tube sqpport plate intersections is based on the analysis contained in WCAP-12871, Revision 2, "J.

M.

Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates," and documentation contained in EPRI Report TR-100407, Revision 1, "PWR Steam Generator Tube Repair TAmi ts - Technical Support Document for Outside Diameter I

Stress Corrosion Cracking at Tube Support Plates. " The application of this i

criteria is based on limiting prinary-to-secondary leakage during a steam line break to lese-than-lwyel4en-pet-minute.

'rimary-t e :::cndc ry 10 kage-dur4ng thfe cycle only i: limit-ed-tc 150 gallen per day-pe r steam-geneteter-dur4*g normal-sperations ensure the applicable Part 100 limits are not exceeded.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

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

Plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness.

If a sleeved tube is found to have through wall penetration of greater than or equal to 31% for the mechanical sleeve and 379 for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R. G.

Changes marked with Gt-rikethrcugho and Bold, Italica 1.121 calculations with 20% added for conservatism. The portion of the tube and the sleeve for which indications of wall degradation must be evaluated can be summarized as follows:

FARLEY-UNIT 2 B 3/4 4-3 AMENDMENT NO. 94 1

P Changes marked with Strikethroughc and Bold, Italics REACTOR COOLANT SYSTEM BASES 3/4.4.7 REACTOR COOLANT SYSTEM LEAYAGE 3/4.4.7.1 LEAKAGE DETECTION SYSTEMS 1

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

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

3/4.4.7.2 OPERATIONAL LEAYAGE 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 seals exceeds 31 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.

1 The surveillance requirements for RCS Pressure Isolation Valves provide l

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 a portion of the allowed limit.

The total steam generator tube leakage limit of 4-GpH d50 gallons per day for all steam generators and 150 gallons per day for any one steam generator 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 1 CP" limit is limits are consistent with the assumptions used in the analysis of these accidents. The 600 250 gpd leakage lindt per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

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

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

EARLEY-UNIT 2 B 3/4 4-4

Changes marked with Strikcthroughc and Bold, Italics i

REACTOR COOLANT SYSTEM r

i BASES 3/4.4.8 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor

{

Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen,

[

chloride and fluoride limits are time and temperature dependent.

Corrosion i

f studies show that operation may be continued with containment concentration levels in excess of the Steady State Limits, up to the Transient Limits, for t

the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the containment concentrations to within the Steady State Limits.

t The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.9 SPECIFIC ACTIVITY t

The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits fel4 ewing a steer generater sube-ruptere-eeefelent-4e-een9 net-len-whF :n ce=umed steady : tete-pr4mery-to--

secondary--eteem-genereter lecheg rate of 1.0 CPM The values for t4: limit en :peci fic ::tivut-represent-14mR: b : d upon a parametri: cvaleetion by the NRC of-ttpleel site location:

These value are-eeneeevetively ir that epeel-f4 sit; p : meter; cf th: F : Icy sit;, such : sit; boundary location end mete relegie:1 condition:, u :: not eeneidered ir thi: ev luctden,-

f 1

For4he--Tenth Ope ret 4eg--Gyel: Only,-the lin Retion: On th: specifi ::tivity l

ef--the-pr4 mary : lent-have been - redu;;d. The : duction in specifie :tivit-y 44mRe-eenbinue t e nsure--t-het--t h ::: ult 4ng 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> de::: st-th: cit: boundary will not ent::d%n appropr4+tely-emal4-f+eet4en-of-pert 100 limR-e in the event of primary-to-secondary leakage as a result of a steam line break.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 4r0 0.Sl microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

i i

FARLEY-UNIT 2 B 3/4 4-5 AMENDMENT NO. 94 F

l

t t

t i

Unit 2 Technical Specification Pages I

i I

l 1

i i

i

e o

?

REACTOR COOLANT SYSTEM i

SURVEILLANCE REQUIREMENTS (Continued) 2 Tubes in those areas where experience has indicated

)

potential problems.

3.

At least 3% of the total number of sleeved tubes in all three steam generators 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.

4.

A tube inspection (pursuant to Specification 4.4.6.4.a 8) shall be performed on each selected tube.

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

L 5.

Tube support plate indications left in service as a result of application of the tube support plate plugging criteria shall be inspected by bobbin coil probe during the following refueling outages.

c.

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:

i 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.

d.

Implementation of the steam generator tube / tube support plate plugging criteria requires 100 percent bobbin coil inspection l

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

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

EARLEY-UNIT 2 3/4 4-10 AMENDMENT NO.

e REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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

4.4.6.2.2 Steam Generator F* Tube Inspection - In addition to the minimum sample size as determined by Specification 4.4.6.2.1, all F*

tubes will be inspected within the tubesheet region.

The results of this inspection will not be a cause for additional inspections per Table 4.4-2.

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

a.

The first inservice inspection 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 n,ot 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-1 category or if two consecutive inspections demonstrate that previously observed degradation 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 l

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 4.4.6.3.a; 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:

EARLEY-UNIT 2 3/4 4-11 AMENDMENT NO.

l

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1.

Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.7.2.

2.

A seismic occurrence greater than the Operating Basis l

Earthquake.

3.

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

4.

A main steam line or feedwater line break.

4.4.6.4 Acceptance Criteria a.

As used in this Specification:

{

l.

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

Eddy-current testing indications below 20% of the l

nominal 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 or sleeve.

3.

Degraded Tube means a tube, including the sleeve if the tube has been repaired, that contains imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

4.

% Degradation means the percentage of the tube or 4

sleeve 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 or sleeve containing a defect is defective.

EARLEY-UNIT 2 3/4 4-12 AMENDMENT NO.

l

-m+

+

4 e

f REACTOR COOLANT SYSTEM P

SURVEILLANCE REQUIREMENTS (Continued) 6.

Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be repaired (i.e.,

sleeved) or removed from service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. This definition does not apply to the area of the tubesheet region below the F* distance in the F*

tubes.

For a tube that has been sleeved with a mechanical joint sleeve, through wall penetration of greater then or equal to 31% of sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by plugging.

For a tube that has been sleeved with a welded joint sleeve, through wall penetration greater than or equal to 37% of sleeve nominal wall thickness in the sleeve between the weld joints requires the tube to be removed from service by plugging. This definition does not apply to tube support plate intersections for which the voltage-based plugging criteria are being applied.

Refer to 4.4.6.4.a.14 for the plugging limit applicable to these 3

intersections.

~

7.

Unserviceable describes the condition of a tube or sleeve if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Badis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.6.3.c, above.

8.

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.

For a tube that has been repaired by sleeving, the tube inspection should include the sleeved portion of the tube.

i 9.

Tube Repair refers to mechanical sleeving, as described i

by Westinghouse report WCAP-11178, R<ev. 1, or laser

)

welded sleeving as described by Westi.,qhouse report WCAP-12672, which is used to maintain a tube in service or return a tube to service.

This inclu.jes the removal of plugs that were installed as a corrective or preventive measure.

FARLEY-UNIT 2 3/4 4-12a AMENDMENT NO.

l

1 g a r

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1 10..

Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to i

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.

11.

F* Distance is the distance of the expanded portion of a tube which provides a sufficient length of undegraded tube expansion to resist pul)out of the tube from the tubesheet.

The F* distance is equal to 1.79 inches and is measured down from the tcp of the tubesheet or the bottom of the roll transition, whichever is lower in elevation.

12.

F* Tube is a tube:

l a) with degradation equal to or greater than 40% below the F* distance, and b) which has no indication of l

imperfections greater than or equal to 20% of nominal wall thickness within the F* distance, and c) that remains inservice.

l 13.

Tube Expansion is that portion of a tube which has been increased in diameter by a rolling process such that no i

crevice exists between the outside diameter of the tube and the hole in the tubesheet.

14.

Tube Support Plate Plugging Limit is used for the disposition of a steam generator tube for continued service that is experiencing outside diameter stress corrosion cracking confined within the thickness of the tube support plates.

These criteria are applicable for l

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

a.

Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltage less than or equal to 2.0 volts will be allowed to remain in service.

b.

Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged except as noted in 4.4.6.4.a.14.c below.

FARLEY-UNIT 2 3/4 4-12b AMENDMENT NO.

i-*

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l c.

Indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to 5.6 volts may remain in service t

if a rotating pancake coil inspection does not detect degradation.

Indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than 5.6 volts will be plugged or repaired, b.

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

l FARLEY-UNIT 2 3/4 4-13 AMENDMENT NO.

)

a

• t l

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l 4.4.6.5 Reports I

a.

Following each inservice inspection of steam generator tubes, I

the number of tubes plugged, repaired or designated F* in each steam generator shall be reported to the Commission within 15 days of the completion of the inspection, plugging or repair effort.

b.

The complete results of the steam generator tube and sleeve 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 and sleeves inspected.

2.

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

l 3.

Identification of tubes plugged or repaired.

c.

Results of steam generator tube inspections which fall into Category C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

d.

For implementation of the voltage-based repair criteria to tube support plate intersections, notify the staff prior to returning the steam generator to service (Mode 4) should any of the following conditions arise:

1.

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

2.

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

3.

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

4.

If the calculated conditional burst probability exceeds 1 x 10'#, notify the NRC and provide an assessment of the safety significance of the occurrence.

FARLEY-UNIT 2 3/4 4-13a AMENDMENT NO.

l

,. g REACTOR COOLANT SYSTEM i

OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be lindted to:

a.

No PRESSURE BOUNDARY LEAKAGE,

]

b.

1 GPM UNIDENTIFIED LEAKAGE, c.

Primary-to-secondary leakage through all steam generators l

shall be limited to 450 gallons per day and 150 gallons per day through any one steam generator.

I d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and 1

e.

31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 i 20 psig, f.

The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve shall be as specified in Table 3.4-1 at a pressure of 2235 i 20 psig.

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 SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1 b.

With any Reactor Coolant System leakage greater than any one j

of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, 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 j

SHUTDOWN within the fellowing 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 limit specified in Table 3.4-1,

)

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 wit.hin 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 fsl.owing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l EARLEY-UNIT 2 3/4 4-17 AMENDMENT NO.

i

o ** %

REACTOR COOLANT SYSTEM 3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.9 The specific activity of the primary coolant shall be limited to a.

Less than or equal to 0.5 ndcrocurie per gram DOSE EQUIVALENT I-131; b.

Less than or equal to 100/E microcurie per gram.

APPLICABILITY:

MODES 1, 2, 3,

4 and 5 ACTION:

MODES 1, 2 and 3*:

a.

With the specific activity of the primary coolant greater than 0.5 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit shown on Figure 3.4-1, be in at least HOT STANDBY with T,yg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With the specific activity of the primary coolant greater l

than 100/E microcurie per gram, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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

avg FARLEY-UNIT 2 3/4 4-23 AMENDMENT NO.

s REACTOR COOLANT SYSTEM ACTION:

(Continued)

MODES 1, 2, 3, 4 and 5:

a.

With the specific activity of the primary coolant greater than 0.5 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E ndcroCuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific i

activity of the primary coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

l FARLEY-UNIT 2 3/4 4-24 AMENDMENT NO.

I

TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE

,q AND ANALYSIS PROGRAM

• D t*

tu TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE 7

AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED s

l C

1.

Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3, 4

l l

N 2.

Isotopic Analysis for DOSE 1 per 14 days 1

l EQUIVALENT I-131 Concentration 3.

Radiochemical for j 1 per 6 months

• 1 Determination 4.

Isotopic Analysis for Iodine a)

Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1#, 2#, 3#, 4#, 5#

Including I-131, 1-133, and 1-135 whenever the specific activity exceeds 0.5 j.

Ci/ gram DOSE EQUIVALENT 1-131 or 200/E pCi/ gram, and w

2 b)

One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a 1,2,3 i

THERMAL POWER change N

exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

1. Until the specific activity of the primary coolant system is restored within its limits.

• Sample to be taken after.a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

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FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Pettent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.5 Ci/ gram Dose Equivalent I-131 F

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FARLEY-UNIT 2 3/4 4-26 AMENDMENT NO.

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REACTOR COOLANT SYSTEM BASES 3/4.4.6 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 maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revi'sion 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 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 manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation

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would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary l leakage - 150 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operational leakage of this magnitude can be readily detected by existing Farley Unit 2 radiation monitors.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which l

the leaking tubes will be located and plugged or repa;. red.

The repair limit for ODSCC at tube support plate intersections is based on the analysis contained in WCAP-12871, Revision 2, "J.

M.

Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates," and documentation contained in EPRI Report TR-100407, Revision 1, "PWR Steam Generator Tube Repair Limits - Technical Support Document for Outside Diameter Stress Corrosion Cracking at Tube Support Plates."

The application of this criteria j

is based on limiting primary-to-secondary leakage during a steam line break to ensure the applicable Part 100 limits are not exceeded.

1 Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness.

If a sleeved tube is found to have through wall penetration of greater than or equal to 31% for the mechanical sleeve and 37% for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R. G.

1.121 calculations with 20% added for conservatism. The portion of the tube and the sleeve for which indications of wall degradation must be evaluated can be summarized as follows:

FARLEY-UNIT 2 B 3/4 4-3 AMENDMENT NO.

s.

• 3 REACTOR COOLANT SYSTEM BASES 3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7.1 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.

3/4.4.7.2 OPERATIONAL LEAKAGE Industry experience has shosm 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 seals exceeds 31 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of i

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.

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 Isolat. ion valves is IDENTIFIED LEAKAGE and will be considered a portion of the allowed limit.

The total steam generator tube leakage limit of 450 gallons per day for all steam generators and 150 gallons per day for any one steam generator ensures that the dosage contribution from the tube leakage will be limited to a small f raction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The limits are consistent with the assumptions used in the analysis of these accidents. The 150 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

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

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

FARLEY-UNIT 2 B 3/* 4-4 AMENDMENT NO.

REACTOR COOLANT SYSTEM BASES 3/4.4.8 CHEMISTRY t

The limitations on Reactor Coolant System chemistry ensure that corrosion of

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the Reactor Coolant System is ndnindzed and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant Syster, over the life of the plant. The associated effects of exceeding the oxygen, i

chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with containment concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the containment concentrations to within the Steady State Limits.

The sarveillance requirementi provide adequate assurance that concentrations in excesu of the limits will be detected in suf ficient time to take corrective action.

3/4.4.9 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits in the event of primary-to-l secondary leakage as a result of a steam line break.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.5 l

microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

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1 FARLEY-UNIT 2 B 3/4 4-5 AMENDMENT NO.

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