ML20046B432

Proposed Tech Specs Modifying TS 3/4.4.6 SG & Bases, TS 3/4.4.7 Reactor Coolant Sys Leakage,Ts 3/4.4.9, Specific Activity & Associated Bases

ML20046B432
Person / Time
Site:	Farley
Issue date:	07/29/1993
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20046B428	List: ML20046B427 ML20046B432
References
NUDOCS 9308040198
Download: ML20046B432 (25)

Text

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f Attachment 3. [

h

.i Revised Unit 2 Technical Specification Pagesf for the- ~i j

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LSteam Generator Tube' Support Plate' Interim' Plugging'Critefia:

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Chanced Pace- Revision.

Page 3/4 4-12  ; Replace ' j

.Page.3/4.4-12a Replace' f Page 3/4 4-17 '

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Page 3/4'.4-23-_ Repl ace . 1 Page 3/4~4-24 -Replace Page'3/4 4-26  ; Replace Page 83/4.'4-3 Replace.

.Page 83/414-3a . Insert.

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[9308040198'930729 PDR -ADOCK '05000364.

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- R'EACTOR C0OLANT SYSTEM-SURVEILLANCE REQUIREMENTS ~(Continued)  ;

j 4.4.6.4 Acceptance Criteria

a. As used in this Specification: o

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

specifications. Ed:fy-current testing indications below 20% of the i nominal wall . thickness, if detectable, may be' considered as ,

imperfections. l

2.

Dearadation means a service-induced cracking,

wastage, wear or 1 general corrosion occurring on either'inside or outside of a tube or sleeve.

3.

Dearaded Tube means a tube,

including Lthe sleeve if the tube has' I been repaired, that contains imperfections. greater'than or equal to

.20% of the. nominal wall thickness caused by degradation. .

4.  % Dearadation 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. . Pluaaina 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:toL40% of the- -

nominal tube wall thickness. This definition does not apply to the l 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, i 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 pl ugging'. At tube support plate intersections, the repair limit for the> Tenth Operating Cycle is based on3 maintaining steam generator l tube serviceability as' described below:

a. An eddy current' examination using a bobbin probe af 100% of the. hot and cold leg steam generator tube support pote

, intersections will..be pe aformed .for- tubes =in. service.

'b. Degradation attributed to outsidi diameterEstress corrosion cracking within the boundsfoi the tube support. plate with  ;

bobbin voltage less than or (qur,' to 1.0 volt'.will be allowed j to-remain in' service. '

1 FARLEY-UNIT 2 '3/4 4-12. AMENDMENT N0.

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

c. Degradation attributed to outside diameter stress corrosion

cracking within the bounds of the tube support plate with a i bobbin voltage greater than 1.0 volt will be repaired or plugged except as noted in 4.4.6.4.a.6.d below.

j d. 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 1.0. volt

but less than or equal to 3.6 volts may remain in service if a rotating pancake coil probe (RPC) inspection does not detect degradation. Indications of outside diameter stress corrosion

- cracking degradation with a bobbin voltage greater than 3.6 volts will be plugged or repaired.

! 7. Unserviceable describes the condition of a tube or sleeve if it

leaks or contains a defect large enough to affect its structural j- . integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as ,

specified in 4.4.6.3.c, above.

8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to i the top support of the cold leg. For a tube that;has be'en 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 Westinghouse report WCAP-)ll78, 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 g

preventive measure.

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

REACTOR COOLANT SYSTEM OPERATIONAL LFAKAGE LIMITING CONDITION FOR OPERATION

, 3.4.7.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE B0UNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE, l

c. For the Tenth Operatina Cycle only. orimary-to-secondary leakaae throuah all steam aenerators shall be limited to 450 aallons per day and 150 aallons per day throuah any one steam aenerator.

, For subsequent cycles,1 GPM total primary-to-secondary leakage l through all steam generators and 500 gallons per day through any one steam generator,

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 i 20 psig.

APPLICABILITY: MODES 1, 2, 3 and 4 l ACTIOJ:

a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT-STANDBY -

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

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT

STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the

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

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

L FARLEY-UNIT 2 3/4 4-17 AMENDMENT N0.

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.- R'EACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS l

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

a. Monitoring the containment atmosphere particulate radioactivity monitor _at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Monitoring the containment air cooler condensate level system or containment atmosphere gaseous radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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

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I REACTOR COOLANT SYSTEM l 3/4.4.9 SPECIFIC ACTIVITY l

l LIMITING CONDITION FOR OPERATION 3.4.9 The specific activity of the 9rimary coolant shall be limited to:

a. Less than or eaual to 0.25 microcurie per aram DOSE E0VIVALENT I-131 for the Tenth Operatina Cycle only:

b. Less than or equal to 1.0 microcurie per gram DOSE EQUIVALENT I-131 for subsequent cycles;

c. Less than or equal to 100/E microcurie per gram. l APPLICABILITY: MODES 1, 2, 3, 4, AND 5 l

ACTION:

MODES 1, 2, AND 3*:

a. For the Tenth Operatina Cycle only. with the specific activity of the primary coolant areater than 0.25 microcurie per aram DOSE EOUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> durina one continuous time interval or

! exceedina the limit line shown on Fiqure 3.4-1. be in at least HOT l STANDBY with T ,, less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

l b. For subsequent cycles, with the specific activity of the primary l l coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for

! more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on. Figure 3.4-1, be in at least HOT STANDBY with T,,, less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With the specific activity of the primary coolant greater than 100/E l microcurie per gram, be in at least HOT STANDBY with T ., less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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

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

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d 4

REACTOR COOLANT SYSTEM ACTION: (Continued)

MODES 1, 2, 3, 4, AND 5

a. For the Tenth Ooeratina Cycle only. with the specific activity of the primary coolant areater than 0.25 microcurie per aram DOSE E0VIVALENT I-131 or areater than 100/E microCuries per aram. Derform the samplina and analysis reauirements of item 4a of Table 4.4-4 until the specific activity of the crimary coolant is restored to within its limits, For subsequent cycles, with the specific activity of.the primary

b. l coolant greater than 1.0 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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' FIGURE 3.4-1 I DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Umit Versus

, Paraant of RATED THERMAL POWER with the Primary Coolant Specific

Activity >1.0pCl/ gram Dose Equivalent 1131 )

{ (Activity >,25p Ci/ gram Dose equivalent I-131 for Cycle 10 only.) .l  !

4 i-3 4

FARLEY-UNIT'2 3/4 4-26 Amendment No ,

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. REACTOR COOLANT SYSTEM ,

BASES

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3/4.4.6 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that-there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion, inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in 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 =

500 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallans per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage _in excess'of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired.

For the Tenth Operating Cycle only, the repair limit for tubes with flaw l indications contained within the bounds of a tube support plate has been provided to the NRC in Southern Nuclear Operating Company letter dated July 29, 1993. The repair limit 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." The application of this criteria is based on limiting primary-to-secondary leakage during a steam line break to less than 1 gallon per minute. Primary-to-secondary leakage during this cycle only is limited to 150 gallons per day per steam generator during normal operation.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness. If a sleeved tube is found to have through wall penetration of greater than or equal to 31% for the mechanical sleeve and 37%

for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R.G. 1.121 calculations with 20% added for conservatism. The portion of the tube and the sleeve for which indications of wall degradation must be evaluated can be summarized as follows:  ;

FARLEY-UNIT 2 B3/4 4-3 AMENDMENT N0.

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R'EACTOR COOLANT SYSTEM BASES

a. Mechanical

1. Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.

2. Indication of tube degradation of any type including a complete guillotine break in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the tube be removed from service.

3. The tube plugging limit continues to apply to the portion of the tube in the entire upper joint region and in the lower roll expansion. As noted above, the sleeve plugging limit applies to these areas also.

4. The tube plugging limit continues to apply to that portion of the tube above the top of the upper joint.

b. Laser Welded

1. Indications of degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.

2. Indication of tube degradation of any type including a complete break in the tube between the upper wcid joint and the lower weld joint does not require that the tee be removed from service.

3. At the weld joint, degradation must be evaluated in both the sleeve and tube.

4. In a joint with more than one weld, the weld closest to the end of the sleeve represents the joint to be inspected.and the limit of the sleeve inspection.

5. The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint.

F* tubes do not have to be plugged or repaired provided the remainder of the tube within the tubesheet that is above the F* distance is not degraded. 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.

Included in this distance is an allowance of 0.25. inch for eddy current elevation measurement uncertainty.

Steam. generator tube inspections ,f operating plants have demonstrated the capability to reliably detect i ige type degradation that has penetrated 20%

of the original tube wall thickne .

FARLEY-UNIT 2 B3/4 4-3a AMENDMENT NO.

REACTOR COOLANT SYSTEM ,

BASES

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Whenever the results of any steam generator tubing inservice inspection fall 1 into Category C-3, these results will be reported to_the Commission pursuant to  !

10 CFR 50.73 prior to resumption of plant operation. Such cases will be ,

considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision to the Technical specifications, if necessary.

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l I-1 FARLEY-UNIT 2 B3/4 4-3b AMENDMENT N0.

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REACTOR COOLANT SYSTEM l BASES l

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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 plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that I 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 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube ruptut ;

accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Farley site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

For the Tenth Operating Cycle only, the limitations on the specific activity of the primary coolant have been reduced. The reduction in specific activity limits continue to ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits in the event of primary-to-secondary leakage as a result of a steam line break.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT l-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.

FA'?iEY- UNIT 2 B 3/4 4-5 AMENDMENT NO.

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Attachment 4 Significant Hazards Consideration ~ Evaluation In Support of .the Technical Specification Changes Associated With Steam Generator Tube Support Plate-Interim Plugging Criteria n

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Joseph M. farley Nuclear Plant - Unit 2 l Technical Specification Changes Associated With Steam Generator Tube Support Plate Interim Plugging Criteria Sianificant Hazards Consideration Analysis i

INTRODUCTION In letters dated January 29, H92 and June 5, 1992, the NRC Staff indicated that they were unable to approve a Technical Specification amendment concerning use of a steam generator tube support plate alternate plugging criteria in time for use in the 1992 outages. On April 1, 1992 and October 8, 1992, the NRC approved 1 volt interim plugging criteria amendments for one cycle of operation of Farley Units 2 and 1. These one cycle amendments are

! not valid beyond the next refueling outages scheduled for the fall of 1993 and spring of 1994 for Units 2 and I respectively. As a result, Southern Nuclear is proposing continued use of the 1 volt interim plugging criteria for use on Farley Unit 2 for the next operating cycle.

DESCRIPTION OF CHANGES As required by 10 CFR 50.91(a)(1), an analysis is provided to demonstrate that the proposed license amendment to implement tne interim plugging criteria for tube support plate elevations involves no significant hazards. The interim plugging criteria involves a correlation between eddy current bobbin probe signal amplitude (voltage) and the tube burst and leakage capability.

Specifically, crack indications with boobin probe voltages less than or equal to 1.0 volt, regardless of indicated depth, do not require remedial action if postulated steam line break leakage can be shown to be acceptable. A sampling program would also be implemented to ensure other forms of degradation are not occurring at Line tube support plates and that cracks are not being masked at I

tube support plates by other factors.

The proposed amendment would modify Technical Specification U4.4.6 " Steam Generators" and its associated bases, Technical Specificatim 3/4.4.7 " Reactor Coolant System Leakage," and Technical Specification 3/4.4.9 " Specific Activity" and its associated bases. The steam generator plugging / 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 I structural integrity. - 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 sp"ecific activity of dose equivalent I"2 and its transient dose equivalent I reactor coolant specific 2

activity is being reduced by a factor of 4.

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. i EVALUATION l

Steam Generator Tube Intearity In the development of the interim plugging criteria, R.G.1.121, " Bases or j Plugging Degraded PWR Steam Generator Tubes," and R.G. 1.83, " Inservice a

Inspection of PWR Steam Generator Tubes,".are used as the bases for i 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 i

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, i should be removed from service. by plugging or repair. Th4s regulatory guide uses safety factors'on loads for tube burst that are consirtent with the requirements Of Section III of the ASME Code. For the tube support plate

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elevation degradation occurring in the Farley steam generators, tube burst

criteria are inherently satisfied during normal operating' conditions by the .

4 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. It is not certain l whether the tube support plate would function to provide a similar j constraining effect during accident condition loadings. Therefore, no credit is taken in the development of the plugging criteria for the presence of the

tube support plate during accident condition loadings. Conservatively, based j on the existing data base, ourst testing shows that the safety requirements

. for tube burst margins during both normal and accident condition loadings can i be satisfied with bobbin coil signal amplitudes several times larger than the proposed 1 0 volt interim plugging criteria, regardless of the depth of tube j 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, j Upon imp'eninntation of the interim plugging criteria, tube leakage i

considerations must also be addressed. It must be determined that the cracks will not leak excessively during all plant conditions. For the interim tube -

plugging criteria developed for the steam generator tuba , no leakage is expected during normal operating conditions even with the presence of through-wall cracks. This is the case as the stress corrosian cracking occurring in the tubes at the support plate elevations in the arley steam generators are short, tight, axially oriented microcracks 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 (SLB) event. For 7/8 inch tubing, the data supports no leakage up to 2.8 volts and a low probability of leakage 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. j l

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me 5dditional 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 amendment is analyzed using the following standards and l 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 I previously evaluated; or 3) involve a significant reduction in a margin of i safety.

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

1) Operation of Farley Unit 2 in accordance wit'1 the proposed license amendment does not involve a sign'fiunt increase in the probability or consequences of an accie'ent previously evaluated.

Testing of model boiler specimens for free standing tubes at room temperature conditions show 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 10 volt indications show burst pressures in excess of 5900 psi at room temperature. Correcting for the effects of temperature on material properties and minimum strength levels (as the burst testing was done at room temperature), tube burst capability significantly exceeds the R.G. 1.121 criterion requiring the maintenance of a margin of three times-normal operating pressure differential on tube burst if through-wall cracks are present.

Based on the existing data base, this criterion is satisfied with bobbin coil indications with signal amplitudes several times the 1.0 volt intarim plugging criteria, regardiess of the indicated depth mease ement. This structural limit is based on'.a lower 95%

confidence level limit of the data. The 1.0 threshold volt

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criteria provides an extremely conservative margin'of safety to the structural . limit considering expected growth rates of ODSCC .

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 3

NNALYSIS (Continued has been ,epsrted from tubes with indications with a voltage level of under 7.7 volts for a 3/4 inch tube with a 10 volt correlation to 7/P inch tubing (as compar)d to the 1.0 volt proposed interim tube plugging 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 conditioa lcauings, the accidents that are affected by primary-to-secondary ledage and steam release to the environment are Lass of External i Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube Rupture, Reactor Coolant Pump Locked Rotor, and ,

Rupture of a Control Roo Drive Mechanism Housing. 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 stea.n release from the secondary side to the environment assume that'tae secondary side remains. intact.

The steam generator tubes are not sutjected to a sustained increase in differential pressure, as is.the case following a  :

steam line break event. Th's increast in. differential pressure is responsible for the postulated increase in leakage and

! associated offsite doses following a steam line break event.

Upun implementation of the interim plugging criteria, it must be '

verified that the expected distribution 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. Date. indicate that a threshold voltage of 2.8 volts would result in through-wall iracks long enough to leak at stb conditions. Application of the proposed plugging criteria reauires that the current distribution of a number of ,

indications versus voltage be obtained during the refueling outages. The current voltage is then combined with the rate of change in voitage measurement to establish an end of cycle voltage distribution and, thus, leak rate during SLB pressure '

di f ferenti al . If it is found that the potential SLB 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 SLB leakage potential to within the acceptance limit. Thus, the consequences of the most limiting des'gn basis accident are l constrained to present licensing bas's limits.

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

Implenentation of the proposed interim tube support plate l elevation steam generator tube plugging criteria does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanirm which 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 plugging r

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. i ANALYSIS (Continued) criteria has been applied (during all plant conditions). The 4 bobbin probe signal amplitude plugging criteria is established such that operational leakage or excessive leakage during a postulated steam line break condition is not anticipated. SNC will implement a maximum leakage rate limit of 150 gpd per steam generatur on Unit 2 to help preclude the potential for excessive leakage during all plant conditions upon application of the plugging criteria. The R.G.1.121 criterion for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture. The 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 operd 5 9 leakage limits are based on leak-before-break consideratim;s 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 three

! against bursting at normal operating pressure differential. A '

voltage amplitude of approximately 4 volts for typical ODSCC 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 .

l through-wall crack length correlations are used below to define the " longest permissible crack" for evaluating operating leakage limits.

1 The single through-wall crack lengths that result in tube burst at three times normal operating pressure differential and SLB conditions are about 0.42 inch and 0.84 inch, respectively.

Normal leakage fer these crack lengths would range from 0.11 '

gallons per minute ta 4.5 gallons per minute, respectively, while lower 95% confidence avel leak rates would range from about 0.02 gallons per minute to 0.5 gallons per minute, respectively.

An operating leak rate of 150 gpd per steam generator will be implemented in application of _ the tube plugging limit. 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%

l confidence level leak rates. Thus, the 150 gpd limit provides ,

l for plant shutdown prior to reaching critical crack lengths for l

[ SLB conditions at leak rates less than a lower 95% confidence '

level and for three times normal operating pressure differential at less than nominal leak rates.

! Based on the above, the : implementation of IPC will not ' create the possibility of a new or different kind of accident from any previously evaluated.

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ANALYSIS (Continued) ]

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3) The proposed license amendment does not involve a significant

! reduction in margin of safety.

' The use of the interim tube support plate elevation plugging criteria is demonstrated to maintain steam generator tube _

4 integrity commensurate with the requirements of R.G. 1.121. R.G. I' j 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 4 consequences of steam generator tube rupture. This is

] accomplished by determining the limiting conditions of 5

degradation of steam generator tubing, as established by 4 inservice inspection, for which tubes with unacceptable cracking

should be removed from service. Upon implementation of the

! criteria,. aen under the worst case conditions, the occurrence of l i ODSCC at the tube support plate elevations is not expected to l 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.

2 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 i tube support plate elevations would . result in minimal, and i acceptable primary to secondary leakage during the event and, i hence, help to demonstrate radiological conditions are less than

an appropriate fraction of the 10 CFR 100 guideline.

I 1 In addressing the combined effects of LOCA +-SSE on the steam ,

generator component (as required by GDC 2), it has been '

j determined that tube collapse may occur in the steam generators j at some plants. This is the case as the tube support plates may

. become deformed as a result of lateral loads at the wedge l supports at the periphery of the plate due to either the LOCA 4

rarefaction wave and/or SSE loadings. Then, the resulting l pressure differential on the deformed tubes may cause some of the j tubes to collapse. j l Additionally, the margin to burst for the tubes using the interim ,

plugging criteria is comparable to that currently provided by l l existing technical specifications.

j There are two issues associated with steam generator tube i collapse. First, the collapse of steam generator tubing reduces 1 the RCS flow area through the tubes. The reduction in flow area

} increases the resistance to flow of steam from the core during a

! LOCA which, in turn, may potentially increase Peak Clad Temperature (PCT). Second, there is a potential the partial

i. through-wall cracks in tubes could progress to through-wall L cracks during tube deformation or collapse.

i 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 Unit 1 and 2

reactor coolant system primary locus and, thus, the probability j of breaks in the primary loop piping is sufficiently low that i

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ANALYSIS (Continued) l I

they need not be considered in the structural design basis of the  !

pl ant. Excluding breaks in the RCS primary loops, the LOCA loads  !

i from the large branch line breaks were analyzed at Farley Unit 1

] and 2 and were found to be of insufficient magnitude to result in steam generator tube collapse or significant deformation.

Regardless of whether or not leak-before-break is applied to the i 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 ii.-

leakage would be increased over current expected levels. Fo" all ',

other steam generator tubes, the possibility of secondary-ta-4 primary leakage in the event of a LOCA + SSE event is not

significant. In actuality, the amount of secondary-to-pr. mary i leakage in the event of a LOCA + SSE is expected to be less than

that currently allowed, i.e., 500 gpd per steam generator.

4 Furthermore, secondary-to-primary in-leakage would be less than primary-to-secondary leakage for the same pressure differential 4 since the cracks would tend to tighten under a secondary-to-primary pressure differential. Also, the presence of the tube l support plate is expected to reduce the amount of in-leakage.

i Addressing the R.G.1.83 considerations, implementation of-the tube plugging criteria is supplemented by 100% inspection requirements at the tube support plate elevations having ODSCC

indications, reduced operating leak rate 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

! principal degradation mechanism as ODSCC.

As noted previously, implementation of the tube support plate elevation plugging criteria will decrease the number of tubes which 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 interim plugging cr!teria will maintain the margin of flow that would otherwise be reduced through increased tube plugging or sleeving. ]

Based on the above,-it is concluded that the proposed change does not result in a significant reduction in margin with respect to i plant safety as defined in the Final Safety Analysis Report or any bases of the plant -Technical Specifications. ,

1 CONCLUSION 4

Based on the preceding analysis, it is concluded that using the interim steam generator tube plugging criterion for remov N tubes from service or repairing tubes at Farley is acceptable and the propm. a license amendment does not involve a-Significant Hazards Consideration nding-as' defined in 10 CFR 50.92.

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Attachment 5 I

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Proposed Eddy Current Guidelines l

For Use With The

! Steam Generator Tube Support-Plate-l ' Interim Plugging Criteria

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Eddy Current Guidelines For Use With The Interim Plugging Criteria This attachment contains guidelines which provide direction in applying the interim plugging criteria. The following items define probe specifications, calibration requirements, specific acquisition and analysis criteria, and flaw recording guidelines to be used for the inspection of the steam generators.

Bobbin Coil Probe

1. Bobbin Coil Probe Specification See Section A.2.1 of Appendix A to WCAP-12871', Revision 2.

2. Bobbin Coil Cal oration Standard See Section A.2.2 and A.2.3 of Appendix A to WCAP-12871, Revision 2.

3. Bobbin Coil Data Acouisitica and Analysis See Section A.2.4, A.2.~ m. ^ 2.6 of Appendix A to WCAP-12871, Revision 2.

Data evaluation of the bobbin signal will be conducted in accordance with Sections A.3.1, A.3.2, A.3.3, A.3.4, and A.3.7 of Appendix A to WCAP-12871, Revision 2, with the exception that the RPC threshold will be reduced to 1.0 volt from 1.5 volts.

4. Bobbin Coil Flaw Recordina Guidelines All flaw signals on the 400/100 mix channel at tube support intersections must be recorded.-

RPC Probe

1. RPC Probe Specification l See Section A.2.1 of Appendix A to WCAP-12871, Revision 2.

i 2. RPC Calibration Standard See Section A.2.2 of Appendix A to WCAP-12871, Revision 2.

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3. RPC Data Acouisition and Analysis j All tube support intersections with bobbin coil flaw indications l registering greater than 1.0 volt shall be inspected with the RPC.

l l See Section A.3.6 of Appendix A to WCAP-12871, Revision 2.

4. RPC Flaw Recordina Guidelines For TSP intersections with a bobbin flaw indication voltage greater than 1 volt, all RPC indications of flaws shall be recorded.

1 Berortina Reauirements l 1. Southern Nuclear will inform the NRC, prior to restart from the refueling l

outage, of any unexpected inspection findings relative to the assumed l characteristics of the flaws at TSPs. This includes any detectable circumferential indications or detectable indications outside the TSP thickness. Any applicable safety evaluations for unexpected findings will be provided to the NRC.

t l 2. The predicted SLB leakage will be reported to the NRC Staff prior to l

restart from the refueling outage.

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. I Attachment 6 Environmental Evaluation In Support of the Technical Specification Changes Associated With Steam Generator Tube Support Plate <

Interim Plugging Criteria l

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I Joseph M. Farley Nuclear Plant - Unit 2 Technical Specification Changes Associated With Steam Generator Tube Support Plate Interim Plugging Criteria Environmental Evaluation Pursuant-to 10CFR51.22(c)(9), the proposed license amendment can be categorically excluded from the requirement to perform an environmental assessment or an environmental impact statement based on the following evaluation:

Southern Nuclear Operating Company has determined that the proposed changes to the Farley technical. specification' associated with the steam generator tube

! support plate interim plugging criteria do not affect the types or amounts of any radiological or non-radiological effluents that may be released offsite.

No increase in individual or cumulative occupational radiation exposure will result from these changes. Additionally, these changes do not involve the use of.any resources not previously considered in the Final Environmental Statement related to the operation of Farley Nuclear Plant.

Based upon this evaluation, it can be concluded pursuant to 10CFR51.22(b) that it is not necessary to perform an environmental assessment or an environmental impact statement.

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