Text

Proposed Tech Specs to Implement voltage-based Repair Criteria for Repair of Westinghouse SG Tubes Affected by Outside Diameter Stress Corrosion Cracking
	ML20077E354
Person / Time
Site:	Farley
Issue date:	12/07/1994
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20077E352	List: ML20077E346; ML20077E354;
References
	NUDOCS 9412120248
	Download: ML20077E354 (71)
	v • d • e

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Attachment 2 Revised Technical Specification Pages Unit 1 Eage 3/4 4-10 Replace 3/4 4-11 Replace 3/4 4-12 Replace 3/4 4-12a Replace 3/4 4-13 Replace ,

3/4 4-13a Insert 3/4 4-23 Replace 3/4 4-24 Replace 3/4 4-26 Replace B3/4 4-3 Replace B3/44-4 Replace B3/4 4-5 Replace Unit 2 Est&C 3/4 4-10 Replace 3/4411 Replace 3/4 4-12 Replace 3/4 4-12a Replace 3/4 4-12b Insert 3/4 4-13 Replace 3/4 4-13a Replace 3/4 4-17 Replace 3/4 4-23 Replace 3/4 4-24 Replace 3/4 4-26 Replace B3/4 4-3 Replace B3/4 4-4 Replace l B3/4 4-5 Replace l

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- REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

1. All nonplugged tubes that previously had detectable wall penetrations greater than 20%.

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

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 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:

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 J imperfections were previously found.

d. Implementation of the steam generator tube / tube support plate plugging criteria requires 100 percent bobbin coil inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 l percent random sampling of tubes inspected over their full length. j The results of each sample inspection shall be classified into one of the following three categories: j I

Catecoty Inspection Results l 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.

r C-3 More than 10% of the total tubes inspected are degraded tubes

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Changes marked with st-r-i-kethroughs and Bold, Itclics or more than 1% of the inspected tubes are defective.

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

FARLEY - UNIT 1 3/4 4-10 AMENDMENT NO.

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Clianges marked with "^'-'reughe and Bold, Italics REACTOR COOLANT SYSTEM f

SURVEILLANCE REQUIREMENTS -(Continued)

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F 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 I service under AVT conditions, not including the preservice l inspection, result in all inspection results falling into the C-1 j category or if two consecutive inspections demonstrate that j 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 l conducted in accordance with Table 4.4-2 at 40 month intervals fall I in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall I apply until the subsequent inspections satisfy the criteria of -)

Specification 4.4.6.3.a; the interval may then be extended to a j maximum of once per 40 months.

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

I following conditions-i

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, j l

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

4. A main steam line or feedwater line break. )

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FARLEY-UNIT 1 3/4 4-11 AMENDMENT NO. l 1

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i j REACTOR COOLANT SYSTEM

. SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.4- Accentance' Criteria !

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a. As used in.this Specification:

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'l. Imperfection means_an exception to the dimensions, finish _or l contour of a tube or sleeve from that required by fabrication -l drawings or specifications. Eddy-current testing indications i below 20% of the 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.

Dearaded 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.

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4. % Dearadation means the percentage of the tube or sleeve wall ;

thickness affected or removed by degradation. -I I

5. Defect means an imperfection of such severity that it exceeds j the plugging or relair limit. A tube or sleeve containing a I defect is defective. I b

6. Pluggina 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' For a tube that has

. l been sleeved with a mechanical joint sleeve, through wall q

l penetration of greater than or equal to 31% of sleeve nominal !

wall thickness in the sleeve requires the tube to be removed l from service by plugging. For a tube that has been. sleeved

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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 i from service by i 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.11 for the plugging limit applicable t.o these intersections. *t tubc cuppset-p-1-are interecctione, the repair limit for t-he-Thieteenth Operating ,

J Cyc? : is haced-en maintaining cteam generatce-bube eerviccability ac-deceribed beleu;

a. 7.n cddy cueront-enamination using a bebbie-prebc cf 1005 ef the het and ccid leg ctcan generator tube cuppe-ob plate interacctienc vill bc performed for tubec'in cervice,

b. Ocgradation attributed to cutside diameter ct-vees I

eerrecien cracking .tithin the bounda cf the tube cupport

Changes marked with Strikethreughs and Bold, Italics i:1000 . tith bebbi: . c i t e,geg g '_' _ _- _4-- - -- . . . alt w: 1 bc 214 ,-._2 ._ __

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FARLEY-UNIT 1 3/4 4-12 AMENDMENT NO.

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' Changes: marked- with Strikethr-eughs and sold, fItalics REACTOR COOLANT SYSTEM i SURVEILLANCE REQUIREMENTS (Continued) '

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c. Ocgradation attri-buted tc cutcide diameter ctrecc escrecicn craching within the bounda of the tub cuppcrt pletc with a bobbir vc1 tag grcater than 2. 0 relt .till be repaired or plugged cncept 2 noted in i i5 d.c.5.d ,

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dy Indicatienc cf pctent-i+' a tradat-ien attributed t-e cutcide diameter strc; cica cracking .ithin the i beund cf the tube cuppa- .. ate .ith a bcbbin vcitage

greater than 2.0 vcit: but lecc than or equal tc 2.5 ve1tc may rcmaic inscrvice if a rotating pancake cci1 prebe (nPC) .incpection dccc net detcet degradation. l Indication: cf cutcide diameter strccc correcien cracking degradation with c bchbin vcitage 7' 'er than 2 . 0 rel t-e will be plugged er repairedy

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 Basis Earthquake, a loss-of-coolant accident, or a steam line or i 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) letely areund the U-bend to the top support of the cold leg or a tube that has been repaired by sleeving, the tube inspection should include the sleeved portion of the tube.

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9. Tube Repair refers to mechanical sleeving, as described by l Westinghouse report WCAP-11178, 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 of plugs that were installed as a corrective or preventive measure.

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' Changes marked with steihwhr-eughs and Bold, Italles EEACTOR 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 techniquea expected to be used during subsequent inservice inspections.

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

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

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 repal:'ed or plugged except as noted in 4. 4. 6. 4. a.11.c below.

c. Indications of potential degradation attributed to outside diameter stress corrosion cracking within the q 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 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.

d. If as a result of leakage due to a mechanism other than outside diameter stress corrosion cracking at the tube support plate intersection, or some other cause, an unscheduled mid-cycle inspection is performed, the following repair criteria apply instead of

4. 4. 6. 4. a .11. c . If bobbin voltage is within expected limits, the indications can remain in service. The expected bobbin voltage limits are determined from the following equation:

A e { V,s - V soc ) + V soc V(CL 1 +( . 2 ) ( A t )

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V= measured voltage, Yaoc = voltage a t BOC, l ht = time period of operation to unscheduled outage, i CL = cycle length (full operating cycle length) where operating cycle is the time between two scheduled steam generator inspections, and Vu= 9.6 volts for 7/8 inch tubes. 1 l

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, 4.4.6.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission within 15 days of the completion of the 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.

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 reaching Mode 2 should any of the following conditions arise:

1. If estimated leakage basce 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.

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3. If indications are identified that extend beyond the confines 1 of the tube support plate.

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4. If the calculated conditional burst probability exceeds 2.5 x 10'#, notify the NRC and provide an assessment of the safety significance of the occurrence.

FARLEY-UNIT 1 3/4 4-13 AMENDMENT NO. l s

'. Changes marked with Strikethroughc and Bold, Italics a

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. Lecc than er equal te 0.25 micrecuric per grcr DOCE EOUIv?.LE r I 121 for the Thirteenth Opcratine cycic cni.

a.h Less than or equal to 4,-O 0.5 microcurie per gram DOSE EQUIVALENT I-131-fer-subecquent cycles; b.er Less than or equal to 100/b microcurie per gram.

APPLICABILITY: MODES 1,.2, 3, 4, and 5 j ACTION:

MODES 1, 2, and 3*:

a. Fcr the Thirteenth Orcrating cycic enl. uitE the cpecific ccticit.

cf the pri= cry ccclant greater th= 0.25 micrcCuric per crcr DOEE ECUIV?.LEm I 121 for mere than de heurc during cnc centinucuc time 4nter-va-1--er eneceding the limit line chc= cn Figurc 2.' 1. he in at

-lea c t HOT ET?2'"nY .ti t h T;...; Iccc than 500 F .ithin C hourc.

ha . Fer cubccquent cyclec, wWith the specific activity of'the primary coolant greater than -1r0 0.5 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hwrs during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,y less than 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , e rb . With the specific activity of the primary coolant greater l than 100/E microcurie per gram, be in at least HOT STANDBY with T,y less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With Tavg greater than or equal to 500*F.

FARLEY-UNIT 1 3/4 4-23 AMENDMENT NO.

Changes marked with Strikethroughc and Bold, Italles 4.

REACTOR COOLANT SYSTEM l

ACTION: (Continued)

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MODES 1, 2, 3, 4, and 5:

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c. "cr the Thirtecnth Operating Cycle enl. .ith the cpecific actifity ef-t-he--_e r ima r-_. reclant creater then 0.25 micrcCuric r_cr c_rc- rocE EOUIVALENT I 131 cr creater than luu/s micrcruricc per grc- perform f the cameline and ancl*/cic requirem^nto cf iter in of Tchic ' ' 4 until the cpecific ccticity cf the primary ccc'lant ic rectored to I within it: limitc

a.br Fee-subsequent cyclec, wWith the specific activity of the primary

coolant greater than -1 r0 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 4 until the specific activity of the primary coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS t

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, f

( FARLEY-UNIT 1 3/4 4-24 AMENDMENT NO.

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PIGURE 3.41 DOSE EQUfVALENT l-131 ".' i Coolam SpecHis Amivity Umit Veum ,

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-u 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.

1 The plant is expected to be operated in a manner such.that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result'in l stress corrosion cracking. The extent of cracking during plant operation would i

be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =

140 gallons per day per steam generator). Cracks having a primary-to-secondary l 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 I accidents. Operational leakage of this magnitude can be readily detected by existing Farley Unit i radiation monitors. 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.

For the Thieteenth Operating Cycle caly, the rcpcir limit for tubec aith fic

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indicatienc contekned aithin the boundo of c tube cupport plate hac beca prccided t4>-the-NRC in Couthrn-Nuclear Operat-ing Ccapany letter dated Decc-ber 00, 1993.

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 is based on limiting primary-to-secondary leakage during a steam line break to 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 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 1 B 3/4 4-3 AMENDMENT NO.

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' Changes marked with Strikethroughs and Bold, Italles REACTOR COOLANT SYSTEM-. !

l BASES 3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7.1 LEAKAGE DETECTION SYSTEMS 1 i

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'l.45, " Reactor Coolant Pressure Boundary Leakage Detection j Systems," May 1973. 1

'l 3/4.4.7.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amountaof 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.

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 LO.'A. . 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 420 gallons per day ,

l 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. steam line break. A-b-GR4 The limit is consistent with the assumptions used in the analysis of these accidents. The 140 gallons per day 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.

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. l l

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FARLEY-UNIT 1 B 3/4 4-4 AMENDMENT NO. )

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' Changes marked with Strikethroughc and Bold, Italics l

l REACTOR COOLANT SYSTEM '

BASES 1

l 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 1 plant. The associated effects of exceeding the oxygen, chloride, and fluoride I limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady ]

State Limits, up to the Transient Limits, for the specified limited time ;

intervals without having a significant effect on the structural integrity of the i Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

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.J SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits fellowing- a cteam generstee--tube rupture acei< lent in conjunction with an accumed-steady ctate--primary te-seeendary steam--generator leakege-rate of 1. 0 CP" The valuce fer-the limito On cpecific I

activity represent limite-baced upon a--paremetric cvaluat4en by the "1C cf t-ypical ci te 1ccat4ena . Thecc valucc are concerva tive--m-t+at cpecific cite paramctcrc of-t-he-Ferlcy citc , cueh cr cite boundary- location and met +ereleg-iea eendi-t4en97 --were-not - concidered in thic cvaluat4enr For-t-he---Thi-rteenth Operating Cycle only, the limitet4 ens-en-enc cpeci-f4e aet4*i+y-of--the--pr-imary-eeelent have been reduced . The reduction in cpecif4e aet-ivit-y-1-imite continuco to ensere-that the result 4ng 0 hour0 days 0 hours 0 weeks 0 months dcccc at t he--eit-e beundary-wi-14-not exceed an appropriately cmall f raction-of--Part 100 limit-e in the event of primary-to-secondary leakage as a result of a steamline break.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 4,-O 0.5 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable l limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

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

l l

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

t

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i w.

l-9 d

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l' u-1 1

Unit 1 Technical specification Pages

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

e e

. REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)-

1. All nonplugged tubes that previously had detectable wall penetrations greater than 20%.

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

3. At'least 3% of the total number of'aleeved tubes in all three steam generators or all of the sleeved tubes in the j generator chosen for the inspection program, whichever is l- 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 i- for a-tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection, i

J

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: -

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 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, i

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

FARLEY-UNIT 1 3/4 4-10 AMENDMENT NO. j l

l i

l e i e .

1

. REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l Category Inapection_.Resulta 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.

4.4.6.3 Inapection 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 inspecticns 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 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 in 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.

FARLEY-UNIT 1 3/4 4-11 AMENDMENT NO.

l

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

0

. 4 ..

.- 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.

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. t 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. Plugaina 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. 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 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.11 for the plugging limit applicable to these 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 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

FARLEY-UNIT 1 3/4 4-12 AMENDMENT NO.

_ _, - _ _ ._ .I

'- REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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.

9. Tube Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-11178, Rev. 1, or laser welded sleeving, as'dcceribed by Westinghouse report WCAp-12672, which is used to maintain a tube in service or return a tube-to service. This includes the removal of plugs that were installed as a corrective or preventive measure.

10. Preservice Innnection 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. 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. 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.11.c below,

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 if a rotating pancake coil '

inspection does not detect degradation. Indications i of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than 5.6 volts will be plugged or repaired.

FARLEY-UNIT 1 3/4 4-12a AMENDMENT NO.

~ -REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

d. If as a_ result of leakage due to a mechanism other than outside diameter stress corrosion cracking at the

- tube support plate intersection, or some other cause, an unscheduled.mid-cycle inspection is performed, the following repair criteria apply instead of 4.4.6.4.a.11.c. If bobbin voltage is within expected-limits, the indications can remain in service. The- .

expected bobbin voltage' limits are determined from the following. equation:

At(V st -Vsoc) + V,oc V(CL 1 + (. . 2 ) ( At)

CL where:

V= measured voltage, Vgg = voltage at-BOC, 1 At= time period of operation to unscheduled outage, CL= cycle length (full operating cycle length) where i operating cycle is the time between two scheduled steam generator inspections, and Vsc = 9.6 volts for 7/8 inch tubes.

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.

4.4.6.5 Reports

a. Following each inservice inspection of' steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the commission within 15 days of the completion of the 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. I

3. Identification of tubes plugged or repaired.

FARLEY-UNIT 1 3/4 4-13 AMENDMENT NO.

e -e i- REACTOR COOLANT' SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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 causelof 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 reaching Mode 2 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 2.5 x 10, notify the NRC and provide an assessment of the safety significance of the occurrence, f

FARLEY-UNIT 1 3/4 4-13a AMENDMENT NO. -l l

1

l l

t REACTOR COOLANT SYSTEM.

3/4.4.9 SPECIFIC ACTIVITY f LIMITING CONDITION FOR OPERATION I l

3.4.9'The specific activity of the primary coolant shall.be limited tot

a. Less than or equal to 0.5 microcurie per gram DOSE EQUIVALENT I-131;

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

I APPLICABILITY: MODES 1, 2, 3, 4, and 5 ACTION: 1 MODES 1, 2, and 3*:

1

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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tn ,, less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the specific activity of the primary coolant greater than 100/E microcurie per gram, be in at least HOT STANDBY with To,,

less than 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

l FARLEY-UNIT 1 3/4 4-23 AMENDMENT NO.

J

a- -e s: REACTOR COOLANT SYSTEM i

ACTION: (Continued) j l

1

~

MODES 1, 2, 3, 4, and 5: l

a. With the specific activity of the primary coolant greater than 0.5 microcurie per gram DOSE EQUIVALENT I-131 or greater than 10 0 / E ;

microcuries per gram, perform the sampling and analysis 1 requirements of item 4a of Table 4.4-4 until the specific activity l of the primary. coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 1

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

PARLEY-UNIT 1 3/4 4-24 AMENDMENT NO.

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O I 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL FOWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.5 pCi/ gram Dose Equivalent I-131 l FARLEY-UNIT 1 3/4 4-26 AMENDMENT NO.

r.. ._ . . . . . . . . . . . - . _ . - . _ . .

1 REACTOR COOLANT SYSTEM

< BASES l l-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 H surveillance of the conditions of the tubes in the event that there is l evidence of mechanical damage or progressive degradation due to design, I manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characte'rizing the nature and cause of any tube degradation so that corrective measures can be taken, The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result'in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 140 gallons per day per steam generator) .

Cracks having a primary-to-secondary leakage less than this limit during l 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 1 radiation monitors. 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.

The repair limit for ODSCC at tube support plate intersections is_ based on the l l

analysis contained in WCAP-12871, Revision S., "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 is based on limiting primary-to-secondary leakage during a steam line break to 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 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 I be summarized as follows: I l

l FARLEY-UNIT 1 B 3/4 4-3 AMENDMENT NO.

_-. - . . . _ ~ ~ . . _ . . ._ _ - . -_ _

d REACTdR 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 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.

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 a portion of the allowed limit.

The total steam generator tube leakage limit of 420 gallons per. day 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 steam line break. The limit is consistent with the assumptions used in the analysis of these ;

accidents. The 140 gallons per day 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.

PRESSURE BOUND.%RY 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.

s

]

l 1

FARLEY-UNIT 1 B 3/4 4-4 AMENDMENT NO. l l

l

4 I

I REACTOR COOLANT SYSTEM. ,

BASES l

1 3/4.4.8 CHEMISTRY j

-1 l

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

Coolant System leakage or failure due to stress corrosion. Maintaining the ;

chemistry within the steady State Limits provides adequate corrosion -

protection to ensure the structural integrity of the. Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits-provides time for taking corrective actions to restore the contaminant concentrations to within the steady State Limits.

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 The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small fraction of Part 200 limits in the event of primary-to- l secondary leakage as a result of a steamline 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.

f f

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

i e - -.. __ -. -- - . . . . . - .

A e

Unit 2 Markups l

a.

Changes marked with Strikethroughs and sold, Italics a.

MEACTOR 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 support plate plugging criteria shall be inspected by bobbin coil probe during the following refueling outages.

t

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:

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 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:

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.

Changes marked with Strikethrougho and Bold, Italics 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 Inapection - 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.

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

l

Changes marked with Strikethroughc and Bold, It:alles REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.3 Inaggetion 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 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 j 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.

l

4. A main steam line or feedwater line break.

FARLEY-UNIT 2 3/4 4-11 AMENDMENT NO. l

. . l Changes; mar W with Strikethroughc and sold, Italics REACTOR COOLANT SYSTEM SURVEILLANCE REQUIRFMENTS (Continued) 4.4.6.4 Aggsptance Criteria a4 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.

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 l gic.Tter than or equal to 20% of the nominal wall t icKness caused by degradation.

4. % Degradation means'the percentage of the tube or i 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. ;

i

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 i of the tubesheet region below the F* distance in.the F*

tubes. For a tube that has been sleeved with a I mechanical joint sleeve, through wall penetration of l 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 greeter 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 culteria are being applied. Refer to

4. 4. 6. 4.a.14 for the plugging limit applicable to these intersections. Ab-bube-eupport pletc interecctienc, j ehc repcir limit fer the Tenth Opercting Cycle is heced j en-mainbeining ctcc genercter tube cerviccchility ce 1 decerihed Scicu; l l

. Changes marked with Strikethroughc and Bold, Italics

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

1 Changes marked with Strikethroughc and sold, ItaIIcs REACTOR COOLANT SYSTEM 1

1

. SURVEILLANCE REQUIREMENTS (Continued) 1 l

c. Degradation attributed te cutsid diameter ctr cc corrccica crc:hing with4n the bound cf the tube cuppcrt plate with = bchbin vcitage grcat=: than 1.0 vclt .till be repnie:d er plugged oncept : i noted ir i 1.C.'. .0,d b 10

d. Indications of potential degradation attr-ibebed tc cutcid; diamet:r str:cc corrccicz craching ,

within the bound cf the tube cuppcrt plate wit-h !

bcbbin vcitage greater ther 1.0 vclt but lecc than er equal to 3.5 volta may remain in cervice if a rotating pancah: coil probe (prC} incpectica dccc not debect degradaticn Indicatienc of outeld: diamatcr ctr cc correcien cracking degradation . tith a bchbir vcitag greater thcn 3.5 vcita vill be plugged er repaired.--

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 Basis Earthquake, a loss-of-coolant accident, = f or a steam.line or feedwater line break as specified in j 4.4,6.3.c, above. )

8. Tube Inspection means an inspection of the steam ,

e 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.

9. Tube Repair re'ers to mechanical sleeving,'as described by Westinghouse report WCAP-11178, 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 of plugs that were installed as a corrective or preventive measure.

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

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 Pluacina Limit is used for the disposition of a stenm generatcr tube for continued service that is experiencing outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate in tersec tions, 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 l support plate with a bobbin voltage greater than 1 2.0 volts will be repaired or plugged except as noted in 4.4.6.4.a.14.c below.

.- . 1 e' 's.

Changes marked with Strikethroughc~and Bold, Italico; .

.i

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 volta may remain in service 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. >

d. If as a result of leakage due to a mechanism '

other than outside diameter stress corrosion cracking at the. tube support plate intersection s or some other cause, an unscheduled mid-cycle inspection is performed, the following repair. .

criteria apply instead of 4.4. 6. 4. a.14. c. If ,

bobbin voltage is within expected limits, the indications'can remain in service. The expecte'd ,

bobbin voltage limits are determined from the following equation:

At'V ss -Vsoc 1 + V soc V{CL 1 +( . 2 ) ( b t )

CL where:

V= measured voltage,

, Vsoc = voltage at BOC, bt = time period of operation to unscheduled outage, _ f CL = cycle length (full operating cycle length) '

where operating cycle is the time between l two scheduled steam generator inspections, and V,s = 9.6 volts for 7/8 inch tubes.

b. The stream 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.

4.4,6.5 Recorts

.j

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged, repaired or designated F* in j each steam generator shall be reported to the Commission .!

within 15 days of the completion of the inspection, plugging l

or repair effort. ;

1

b. The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in 1

Changes marked with Strikethroughc and Bold, Italles 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, i l

l

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

3. Identification of tubes plugged or repaired.

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

I

.1 j

Changes marked with Strikethrougho and Bold, Italics l

s. ;

REACTOR COOLANT SYSTEM I

SURVEILLANCE REQUIREMENTS (Continued) !

l 1

.I

c. Results of steam generator tube inspections which fall into l Category C-3 shall be considered a REPORTABLE EVENT and shall )

be reported pursuant to 10CFR50.73 prior to resumption of l plant operation. The written report shall provide a 1 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 reaching Mode 2 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 2.5 x 10*#, notify the NRC and provide an assessment of the safety significance of the occurrence.

I 1

l l

1 1

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

I 1

i l

I

e Changes' marked with striket-breughe and sold, Italics ,

REACTOR COOLfMr SYSTEM 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. For the T nth Operating Cyclc 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-eubacquent cycicc, 1 C"" total primary tc cccondney leakagc_through all ctcc generaterc and 500 gallenc per day thrcugh any One etcar scncrateFr

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

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

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 1 20 psig.

APPLICA3ILITY: MODES.1, 2, 3 and 4 l

ACTION:

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

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant. System Pressure Isolation Valve leakage greater than the limit specified in Table 3.4-1, .j isolate the high pressure portion of the affected system from j the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , i

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

_ _ __ __ ~_

1 1

I Changes marked with strikethroughs and Bold, Italics !

REACTOR COOLANT SYSTEM ;

3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.3 The specific activity of the primary coolant shall be limited to:

c. Lecc than er ecual to 0.25 micrcCuric per gran DOCE EOUIVALE"T I 131 fer the Tenth Occrctinc Cycle onl, I bra. Less than or equal to 4,4 0.5 microcurie per gram DOSE EQUIVALENT I-131 for cubccquent cyclec; erb. Less than or equal to 100/5 microcurie per gram. I APPLICABILITY: MODES 1, 2, 3, 4 and 5 ACTION: I 1

MODES 1, 2 and 3 * : l

c. O Lthe Tenth Occrcting Cycle onl, .-tith the cpecific activit_, cf 'hc primary ccclant grecter than 0.25 micrcCuric ocr cram MCE EOUIVALE""' I 121 for = crc than 'S heurc during en ct+nt inucuc time interval cr cncceding the limit line shewn On Ficurc ? ' 1. be in at lecct "OT CT?dOEY .Jith T-.,.~

-n lecc thcMGGP 2-tithin C hours. j bra. For-subsequent cyc1cc, w With the specific activity of the primary coolant greater than +r0 0.5 microcurie per gram DOSE ;

EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous I

time interval or exceeding the limit shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

erb. 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i i

With Tavg greater than or equal to 500 F.

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

Changes marked with Strikethroughc.and Bold, Italics REACTOR COOLANT SYSTEM-ACTION: (Continued)

MODES 1, 2, 3, 4 and 5:

3. "cr the Tenth Operating Cvcic ent_. with the cpecific'actifity e4-the-primary cociant greater than ^25 micrcCuric per gram nn e. e e,ni m ,se m - , ,,, _._____-._ut_. , n n ie __e...2__ ___

grac. cerform the ccmpling and analycic requiremento cf item da cf Tchir i e i until the cpecific acti eit-_. cf the primary ecciant in rectered te withic itc limitc.

bra. Fer ubccquent cycles, w With the specific activity of the-primary coolant greater than -1ro 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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l 1

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

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PIGURE 3.41 DOSE EQUNALENT l -131 Primary Coolant SpeeHis AssMey Umit Verum Parenne of THERMAL POWER with the M._i Cooient Specifle Activity > pam Does Estuivalent M31 m..,,c.s m.g ,. .,.,,, __ m_

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FARLEY-UNIT 2 3/4 4-26 i

Amendment N). 94 r

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. .- 1 Changes marked -with st-skee-hrswJh s and _ Bold, Italics

~

] \

REACTOR COOLANT SYSTEM BASES 3/4.4.6 STEAM GENERATORS l

1 The Surveillance Requirements for inspection of the steam generator tubes j ensure that the structural integrity of thia portion of the RCS will be maintained. 'The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide.l.83, Revision 1. Inservice ,

inspection of steam generator tubing is essential in order to maintain j 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 corrcaion.

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. )

1 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 l corrosion of the steam generator tubes. If the secondary coolant chemistry is ;

not maintained within these limits, localized corrosion may likely result in ,

I stress corrosion cracking. The extent of cracking.during plant operation would be limited by the limitation of steam generator tube leakage between the i primary coolant system and the secondary coolant system leakage - 604 150 gallons per day per steam generator). (primary-to-secondary Cracks having a l primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during-normal position and by postulated accidents. Operating plants have demenctented that prenary tc cccendary leakage cf 500 gallenc per day per cteam gencr-at-ee Operational leakage of this magnitude can be readily he detected by existing Farley Unit 2 radiation monitors ef-stccm generater bicudcur. 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.

Fee-t-he---Tenth Operating Cycle caly, t The repair limit for tubec . tith-f4ew indi-eat 4ene--eentwined-within the houndo cf a tub: cuppert plate-hac been previded--tc the imC in Ceuthern Muclear Operst4ng Ocmpany 10tter dat-ed-Jul-y 2D,1993. 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 lo "PNR Steam Generator Tube Repair Limits - Technical Support Document for Outside Diameter Stress Corrosion Cracking at Tube Support Plates." The application of this criteria is based on limiting primary-to-secondary leakage during'a steam line break to 4eee--than 1 gallen per minute. Primary tc cccendary leahag during thie-eyele-enly ic limited tc 150 gallcnc per day per etcar gencret-er during normal apcrat4en,-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 eteam 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 sJeeve nominal wall thickness in l

i

Changes marked with st-rikethr-eughs and Bold, Italles 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.

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i 9

XChanges marked with Strikethroughc and Bold, Italics REACTOR COOLANT SYSTEM BASES 3/4.4.? REACTOR COOLANT SYSTEM LEAFAGE

-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 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 avstems.

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.

The surveillance requiremento 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 a portion of the allowed limit.

The total steam generator tube leakage limit of 4-GPM 450 gallons per day for all steam generators and 130 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 4-GPM limit is consistent with the assumptions used in the analysis of these accidents. The GOO 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.

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 4-4 AMENDMENT NO.

E.

Changes marked with Strikesh mughs and Bold, Italles REACTOR COOLANT SYSTEM BASES 3/414.8' 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 corrosj on. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant, The associated effects'of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion-studies show that operation may be continued with containment concentration levels in excess of the Sct.3dy State Limits, up to the Transient Limits, for-the specified limited time i.itervals 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 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 The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits follc.eing a ates= 3enereber tube-rupter accident-i-n conjunet4cn ,rith an accumed etcady ctate p:1 mary to--

seeendary-eteam-generator-4eakage-rate of 1. ^ CP" The valucc for the limi-se en cpecific activity represent--14mu-c -based - upon a parametric evaluabien-by the ""O cf-bypical c he-locatienc . Thece valuce arc concervabively in t-had opeei-f4e-she--parametees-of the "aeley cite, cuch ac cit: bcundary-leent4en ,

and meteorelegical-eendh4 enc, .: crc not canoidered--in---bh-ie-evaluabien r Fer-the-Tenth-Operat4ng Cycic caly, t he limitat-ienc On - the cpecif4e-aet4viby l o f-the-prima ry-eoolen t-have -- b e e n r e duce d . The-reduetien :: cpecific 'act4+ky l 14mibe-eent4nue to ensure-that - the recuh4ng 2 hour-desee-at the cite boundar-y wi44-net-eneeed-en-appropr4ade4y-sma14-freet4en-cf " art-+00-14*ite in the event of primary-to-secondary leakage as a result of a steam line break. I The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 4-O 0.5l 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, l FARLEY-UNIT 2 B 3/4 4-5 AMENDMENT NO.

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Unit 2 Technical Specification-Pages-

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"+ 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 stear. generators or all of the sleeved tubes in p .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

i. 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 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:

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

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

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

' REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Category InSDECtion Resulta 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 i 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 innpection 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 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 1

performed on each steam generator in accordance with the first sample inspection specified in Table 4.1-2 during the shutdown subsequent to any of the following conditions:

i FARLEY-UNIT 2 3/4 4-11 AMENDMENT NO. l

. s :. e

.. 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 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:

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.

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

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

.s: s REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (C'ontinued) 1-

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 wal1~ 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 thatLhas 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

} 4.4.644.a.14 for the plugging limit applicable to these intersections.

7. Unserviceabla 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 Basis 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 InsDection 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.

9. Tube RRpair refers to mechanical sleeving, as described by Westinghouse report WCAP-11178, 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 of plugs that were installed as a corrective or preventive measure.

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

+

s s

6 - 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 tuber 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 for the disposition of a steam generator tube for continued service that is experiencing outside diameter' stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

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.

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REACTOR COOLANT SYSTEM l.

SURVEILLANCE REQUIREMENTS (Continued) ._

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 if a rotating pancake coil inspection does not t

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

, d. If as a result of leakage due to a mechanism L other than outside diameter stress corrosion cracking at the tube support plate intersection, or some other cause, an unscheduled mid-cycle inspection is performed, the following repair

criteria apply instead of 4.4.6.4.a.14.c. If bobbin voltage is within expected limits, the indications can remain in service. The expected bobbin voltage limits are determined from the following equation:

l l

I At(V st - V ,oc) + V ,oc i.

y(CL j I

1 + ( . 2 ) ( At) i CL where:

V= measured voltage, ),

Vsoc= voltage at BOC, At = time period of operation to unscheduled outage, CL= cycle length (full operating cycle length) where l operating cycle is the time between two scheduled steam generator inspections, and V33 = 9.6 volts for 7/8 inch tubes.

b. The stream 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 l

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

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REACTOR COOLANT SYSTEM.

SURVEILLANCE REQUIREMENTS (Continued)

~

4.4.6.5 Reports

a. Following each inservice inspection of steam generator tubes, 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 Comminsion in a Spacial 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.

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 reaching Mode 2 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 l licensing basis assumptions) during the previous i 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 l confines of the tube support plate. I

4. If the calculated conditional burst probability exceeds 2.5 x 10'*, notify the NRC and provide an assessment of the safety significance of the occurrence, l

)

l l

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

i l

1

- - -. - r

. ,. - . -~ - .

3.

REACTOR COOLANT SYSTEM' OPERATIO'NAL 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. 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, l'

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

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 1 20 psig.

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

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4. hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I 1

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

1 I

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, 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 microcurie-per gram DOSE EQUIVALENT i 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With the specific activity of the primary coolant greater l '

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

4 i

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With Tavg greater than or equal to 500*F.

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

i

. . . ~ _,

l

~*'

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 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 .

l 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 l

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

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20 30 40 50 60 70 80 90 100 PERCEtTr OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific l Activity > 0.5 pCi/ gram Dose Equivalent I-131 FARLEY-UNIT 2 3/4 4-26 AMENDMENT N3.

1 1

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. ' '* t..

~ REACTOR COOIANT 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, 4 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

, would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system leakage = 150 gallons per day per steam generator) . (primary-to-secondary Cracks having a primary- l to-secondary leakage less than this limit during operation will have an

adequate margin of safety to withstand the loads imposed during normal position 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 the leaking tubes will be located and plugged or repaired.

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 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 l sleeve and 37% for the laser welded sleeve of sleeve nominal wall thickness in l the sleeve, it must be plugged. The 31% and 37% limits are derived-from R. G. i 1.121 calculations with 20% added for conservatism. The portion of the tube l

and the sleeve for which indications of wall degradation must be evaluated can be summarized as follows: l l

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

1

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

P 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 OPEPJLTIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is j 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 -I 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.

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 a portion of the I allowed limit.

l The total steam generator tube leakage limit of 450 gallons per day for l 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 stean generator tube rupture or r line break. .The limit is consistent with the ,

assumptions use-I in + ie analysis of these accidents. The 150 gpd leakage l limit per steam generitor ensures that steam generator tube integrity is. l maintained in the v#nt of a main steam line rupture or under LOCA condrtions.

PRESSURE ROUNDARY 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 prongtly placed in COLD SHUTDOWN.

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

w - *-

~ PEACTOR COOLANT' SYSTEM 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 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 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 ACTIVIIY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an 1 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 limi' tu 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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FARLEY-UNIT 2 B 3/4 4-5 AMENDMENT NO, i 1

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Attachment 3 r

Significant flazards Evaluation Voltage-Based Repair Criteria l

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Joseph M. Farley Nuclear Plant

. Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking Sienificant llazards Consideration Analysis j i

l DESCRIPTION OF CIIANGES As required by 10 CFR 50.91(a)(1), an analysis is provided to demonstrate that the proposed license ,

amendment to implement the voltage-based repair criteria for tube support plate elevations involves no )

significant hazards. The voltage-based repair criteria involve a correlation between eddy current bobbin probe signal amplitudes (voltage) and the tube burst and leakage capabilities.

Specifically, crack indications with bobbin probe voltages less than or equal to 2.0 volts, regardless of indicated depth, do not require remedial action if postulated steam line break leakage can be shown to be i acceptable. A sampling program would also be implemented to ensure other forms of degradation are i not occurring at the tube support plates and that cracks are not being masked at tube support plates by other factors.

The proposed amendment would modify Technical Specification 3/4.4.6 " Steam Generators" and its associated bases, Technical Specification 3/4.4.7 " Operational Leakage" and its associated bases for Unit 2, and Technical Specification 3/4.4.9 " Specific Activity" and its associated bases. The steam generator repair limit will be modified to clarify that the appropriate method for determining serviceability for tubes with outside diameter stress corrosion cracking at the tube support plate is by a methodology that more reliably assesses structural integrity. For Unit I, the operational leakage requirement has previously been modified to reduce the total allowable primary-to-secondary leakage for any one steam generator from 500 gallons per day to 140 gallons per day. For Unit 2, the operational leakage requirement will be modified to reduce the total allowable primary-to-secondary leakage for any one steam generator from 500 gallons per day to 150 gallons per day. In addition, the technical specification limit for specific activity of dose equivalent I* and its transient dose equivalent I"' reactor coolant specific activity is being reduced by a factor of 2 in order to increase the allowable leakage in the event of a steam line break.

EVALUATION Sleam Generator Tube Integrity in the development of the voltage-based repair criteria, 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 uwd as the bases for determining that steam generator tube integrity considerations are maintained within acceptable limits. R.G.1.121 describes a method acceptable to the NRC staff for meeting General Design Criteria 2,14,15,31, 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 or repair. This regulatory guide uses safety factors on loads for tube burst that are consistent with the requirements of Section 111 of the ASME Code. For the tube support plate elevation degradation l

e Significant Ilazards Evaluation Page 2 l Voltage-Based Repair Criteria occurring in the Farley 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 of the drilled hole. Analyses in WCAP-12871 show that for open cervices with as-designed gaps, the tube support plate may not function to provide a similar constraining effect during accident condition loadings. The WCAP-12871 analyses for Farley Unit I with corroded and packed crevices, as confirmed by bobbin coil inspection, show that the tube support plates would not be significantly displaced even under steam line break loading conditions. For conservatism in the voltage-based plugging criteria repair criteria, no credit is taken in the development of the repair criteria for the presence of the tube support plate during accident condition loadings. Conservatively, based on the existing data base, burst testing shows that the safety requirements for tube burst margins during accident I condition loadings can be satisfied with bobbin coil signal amplitudes several times larger than the proposed 2.0 volt voltage-based repair criteria, regardless of the depth of tube wall penetration of the cracking. R.G.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.

Upon implementation of the voltage-based repair 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 voltage-based tube repair criteria developed for the steam generator tubes, no leakage is expected during normal operating conditions even with the presence of through-wall cracks. This is the case as the stress corrosion cracking occurring in the tubes at the support plate elevations in the Farley steam generators is short, tight, axially oriented micro cracks ollen separated by ligaments of material. No leakage during normal operating conditions has been observed in the field for crack indications with signal amplitudes less than 7.7 volts in a 3/4 inch tube. Voltage correlation to 7/8 inch tubing size would result in an expected voltage of about 10 volts. Relative to the expected leakage during accident condition loadings, the limiting event with respect to primary-to-secondary leakage is a postulated steam line break event. For 7/8 inch tubing, the data supports no leakage up to 2.8 volts and a low probability 4

ofleakage between 2.8 and 6.0 volts. The threshold of significant leakage (20.31/ hour or 10 gpm) in a 7/8 inch tube diameter is 6 volts.

Additional Considerations The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by plant workers involved in tube plugging or repair operations. The proposed amendment would minimize the loss of margin in the reactor coolant flow through the steam generator by keeping structurally sound tubes in service and not unnecessarily plugging or sleeving them. The proposed amendment would avoid loss of margin in reactor coolant system flow and, therefore, assist in demonstrating that minimum flow rates are maintained in excess of that required for operation at full power. Reduction in the amount of tube plugging and 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.

ANALYSIS In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license i amendment is analyzed using the following standards and found not to: 1) involve a significant increase in the probability or consequences for an accident previously evaluated; or 2) create the possibility of a !

new or different kind of accident from any accident previously evaluated; or 3) involve a significant i reduction in a margin of safety. l l

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' Significant Hazards Evaluation Page 3 Voltage-Based Repair Criteria I l

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

1) Operation of Farley units in accordance with the proposed license amendment does not involve a significant increase in the probability er consequences of an accident previously evaluated.

Testing of model boiler specimens for free standing tubes at room temperature conditions shows burst pressures as high as approximately 5000 psi for indications of outer diameter stress corrosion cracking with voltage measurements as high as 26.5 volts. Burst testing performed on pulled tubes with up to 7.5 volt indications show burst pressures in excess of 5900 psi at room temperature. As stated earlier, tube burst criteria are inherently satisfied during norm'al operating conditions by the presence of the tube support plate. Furthermore, correcting for the effects of I

temperature on material properties and minimum strength levels (as the burst testing was done at room temperature), tube burrt capability significantly exceeds the R.G.1.121 criterion requiring the maintenance of a margin of 1.43 times the steam line break pressure differential on tube burst if through-wall cracks are present without regard to the presence of the tube support plate.

Considering the existing data base, this criterion is satisfied with bobbin coil indications with signal amplitudes over twice the 2.0 volt voltage-based repair criteria, regardless of the indicated depth measurement. This structural limit is based on a lower 95% confidence level limit of the data. The 2.0 volt criterion provides an extremely conservative margin of safety to the structural limit considering expected growth rates of outside diameter stress corrosion cracking at Farley.

Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by a burst pressure to voltage correlation. However, relative to expected leakage during normal operating conditions, no field leakage has been reported from tubes with indications with a voltage level of under 7.7 volts for a 3/4 inch tube which correlates to 10 volts for 7/8 inch tubing (as comparcJ to the 2.0 volt proposed voltage-based tube repair limit). Thus, the proposed amendment does not involve a significant increase in the probability or consequences of an accident.

Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary-to-secondary leakage and steam release to the environment are Loss of External Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Secondary System Pipe Failure, Steam Generatot Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism llousing. Of these, the Major Secondary System Pipe Failure is the most limiting for Farley in considering the potential for off-site doses. The offsite dose analyses for the other events which model primary-to-secondary leakage and steam releases from the secondary side to the environment assume that the secondary side remains intact. The steam generator tubes are not subjected to a sustained increase in differential pressure, as is the case following a steam line break event. This increase in differential pressure is responsible for the postulated increase in leakage and associated offsite doses following a steam line break event. In addition, the steam line break event results in a bypass of containment for steam generator leakage. Upon implementation of the voltage-based repais criteria, it must be verified that the expected distributions of cracking indications at the tube support plate intersections are such that primary-to-secondary leakage would result in site boundary dose within the current licensing basis. Data indicate that a threshold voltage of 2.8 volts could result in through-wall cracks long enough to leak at steam line break conditions.

Application of the proposed repair criteria requires that the current distribution of a number of

- < e

, Significant Hazards Evaluation Page 4 Voltage-Based Repair Criteria indications versus voltage be obtained during the refueling outages. The current voltage is then combined with the rate of change in voltage measurement and a voltage measurement uncertainty to establish an end of cycle voltage distribution and, thus, leak rate during steam line break pressure differential. The leak rate during a steam line break is further increased by a factor related to the probability of detection of the flaws. Ifit is found that the potential steam line break leakage for degraded intersections planned to be left in service coupled with the reduced specific activity levels allowed result in radiological consequences outside the current licensing basis, then additional tubes will be plugged or repaired to reduce steam line break leakage potential to within the acceptance limit. Thus, the con. sequences of the most limiting design basis accident are constrained to present licensing basis limits.

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

Implementation of the proposed voltage-based tube support plate elevation steam generator tube repair criteria does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism that could result in an accident outside of the region of the tube support plate elevations. Neither a single or multiple tube rupture event would be expected in a steam generator in which the repair criteria have been applied during all plant conditions.

The bobbin probe signal amplitude repair criteria are established such that operational leakage or excessive leakage during a postulated steam line break condition is not anticipated. Southern Nuclear has previously implemented a maximum leakage limit of 140/150 gpd (Unit 1/ Unit 2) per steam generator. The R.G.1.121 criterion for establishing operational leakage limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture. The 140/150 gpd limit provides for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest permissible crack length. R.G.1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded. The longest permissible crack is the length that provides a factor of safety of 1.43 against bursting at steam line break pressure differential. A voltage amplitude of approximately 9 volts for typical outside diameter stress corrosion cracking corresponds to meeting this tube burst requirement at the 95% prediction interval on the burst correlation. Alternate crack morphologies can correspond to a voltage 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 is used below to define the " longest permissible crack" for i evaluating operating leakage limits. l The single through-wall crack lengths that result in tube burst at 1,43 times steam line break pressure differential and steam line break conditions are about 0.53 inch and 0.84 inch, respectively. Normal leakage for these crack lengths would range from about 0.4 gallons per minute to 4.5 gallons per minute, respectively, while lower 95% confidence level leak rates ]

would range from about 0.06 gallons per minute to 0.6 gallons per minute, respectively. l An operating leak rate of 140/150 gpd per steam generator has been implemented. This leakage limit provides 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 140/150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for steam line break conditions at leak rates less l

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Significant Hazards Evaluation Page 5 l Voltage Based Repair Criteria ;

than a lower 95% conGdence level and for three times normal operating pressure differential at less than nominal leak rates.

l Considering the above, the implementation of voltage-based plugging criteria will not create the possibility of a new or different kind of accident from any previously evaluated.

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

The use of the voltage-based tube support plate elesation repair criteria is demonstrated to maintain steam generator tube integrity commensurate with the requirements of R.G.1.121.

R.G.1.121 describes a method acceptable to the NRC staff for meeting GDCs 2,14,15,31, and 32 by reducing the probability of the consequences of steam generator tube rupture. This is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established 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 outside diameter stress corrosion cracking 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 most limiting effect would be a possible increase in leakage during a steam line break event. Excessive leakage during a steam line break event, however, is precluded by verifying that, once the criteria are applied, the expected end of cycle distribution of crack indications at the tube support plate elevations would result in minimal, and acceptable primary to secondary leakage during the event and, hence, help to demonstrate radiological conditions are less than an appropriate fraction of the 10 CFR 100 guideline.

The margin to burst for the tubes using the voltage-based repair criteria is comparable to that currently provided by existing technical speci6 cations.

In addressing the combined effects of LOCA + SSE on the steam generator component (as required by GDC 2), it has been determined that tube collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become deformed as a result of lateral loads at the wedge supports at the periphery of the plate due to either the LOCA rarefaction wave and/or SSE loadings. Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.

There are two issues associated with steam generator tube collapse. First, the collapse of steam generator tubing reduces the RCS How area through the tubes. The reduction in How area increases the resistance to Dow of steam from the core during a LOCA which, in turn, may potentially increase Peak Clad Temperature (PCT). Second, there is a potential the partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse or that short through-wall indications would leak at significantly higher leak rates than included in the leak rate assessments.

Consequently, a detailed leak-before-break analysis was performed and it was concluded that the leak-before-break methodology (as permitted by GDC 4) is applicable to the Farley reactor coolant system primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design basis of the plant.

Excluding breaks in the RCS primary loops, the LOCA loads from the large branch line breaks i

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o i Significant flazards Evaluation Page 6 Voltage-Based Repair Criteria .

\ l were analyzed at Farley and were found to be ofinsufficient magnitude to result in steam i generator tube collapse or significant deformation. j Regardless of whether or not leak-before-break is applied to the primary loop piping at Farley, any flow area reduction is expected to be minimal (much less than 1%) and PCT margin is available to account for this potential effect. Based on analyses' results, no tubes near wedge locations are expected to collapse or deform to the degree that secondary to primary in-leakage would be increased over current expected levels. 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 amount of secondary-to-primary leakage in the event of a LOCA +

SSE is expected to be less than that previously allowed, i.e.,500 gpd per steam generator.

Furthermore, secondary-to-primary in-leakage would be less than primary-to-secondary leakage for the same pressure differential since the cracks would tend to tighten under a secondary-to-primary pressure differential. Also, the presence of the tube support plate is expected to reduce the amount of in-leakage.

Addressing the R.G.1.83 considerations, implementation of the tube repair criteria is supplemented by 100% inspection requirements at the tube support plate elevaticas having outside diameter stress corrosion cracking indications, reduced operating leakage limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotating pancake coil inspection requirements for the larger indications left in service to characterize the principle degradation mechanism as outside diameter stress corrosion cracking.

As noted previously, implementation of the tube support plate elevation repair criteria will decrease the number of tubes that must be taken out of service with tube plugs or repaired. The installation of steam generator tube plugs or tube sleeves would reduce the RCS flow margin, thus implementation of the voltage-based repair criteria will maintain the margin of flow that would otherwise be reduced through increased tube plugging or sleeving.

Considering the above, it is concluded that the proposed change 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.

CDECLUSION Based on the preceding analysis, it is concluded that using the voltage-based steam generator tube repair criterion for removing tubes from service or repairing tubes at Farley is acceptable and the proposed license amendment does not involve a Significant llazards Consideration as defined in 10 CFR 50.92.

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ML20077E354

Text

Dearaded Tube means a tube,

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