ML20058J567

Proposed Tech Specs Associated W/Steam Generator Tube Support Plate Interim Repair Criteria

ML20058J567
Person / Time
Site:	Farley
Issue date:	12/09/1993
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20058J557	List: ML20058J553 ML20058J567
References
NUDOCS 9312140158
Download: ML20058J567 (17)

Text

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Attachment 3

, Revised Unit 1 Technical Specification Pages forthe Steam Generator Tube Support Plate Interim Repair Criteria Chanaed Page , Revision Page 3/4 4-12 Replace Page 3/4 4-12a Replace Page 3/4 4-17 Replace Page 3/4 4-17a insert Page 3/4 4-23 Replace Page 3/4 4-24 Replace Page 3/4 4-26 Replace Page B3/4 4-3 Replace Page B3/4 4-3a insert Page B3/4 4-3b insert Page B3/4 4-5 Replace l

9312140158 931209 ,

PDR ADOCK 05000348 l p PDR ,

REACTOR COOLANT SYSTEM SURVEILIANCE REQUIREMENTS (Continued)

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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 01 sleeve trom 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 i general corrosion occurring on either inside or outside of a tube  :

or sleeve.

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

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4.  % Degradation means the percentage of the tube or sleeve wall l 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. [

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6. Plugging or Repair Limit means the imperfection depth at or beyond  !

which the tube shall be repaired (i . e . , sleeved) or removed from i service by plugging and is greater than or equal to 40% of the I nominal tube wall thickness. For a tube that has been sleeved f with a mechanical joint sleeve, through wall penetration of l greater than or equal to 31e of sleeve nominal wall thickness in l 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 l sleeve nominal wall thickness in the sleeve between the weld '

joints requires the tube to be removed from service by plugging. l At tube support plate intersections, the repair limit for the l

Thirteenth Operating Cycle is based on maintaining steam l generator tube serviceability as described below:

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a. An eddy current examination using a bobbin probe of-100% of'  ;

, the hot and cold leg steam generator tube support plate l intersections will be performed for tubes in service. .;

b. Degradation attributed to outside diameter stress corrosion ~

cracking within the bounds of the tube support plate with l bobbin voltage less than or equal to 2.0 volts will be l  !

allowed to remain in service.

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l EARLEY-UNIT 1 3/4 4-12 AMENDMENT NO. [

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l REACTOR COOLANT SYSTEM -l 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 2.0 volts will be repaired or l plugged except as noted in 4.4.6.4.a.6.d below. l

d. Indications of potential degradation attributed to outside f diameter stress corrosion cracking within the. bounds of the -

tube support plate with a bobbin voltage greater than 2 O  ;

volts but less than or equal to 5.6 volts may remain in i service if a rotating pancake coil probe (RPC) inspection i does not detect degradation. Indications of outside  !

diameter stress corrosion cracking' degradation with a bobbin l voltage greater than 5.6 volts will be plugged or repaired. l i

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

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

specified in 4.4.6.3.c, above.

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8. Tube Inspection means an inspection of the steam generator tube l 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 j sleeved portion of the tube.

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9. Tube Repair refers to mechanical sleeving, as described by [

Westinghouse report WCAP-11178, Rev. 1, or laser welded sleeving, j 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. L{

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EARLEY-UNIT 1 3/4 4-12a AMENDMENT NO.  !

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REACTOR COOLANT SYSTEM 3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION

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3.4.9 The specific activity of the primary coolant shall be limited to:

a. Less than or equal to 0.25 microcurie per gram DOSE EQUIVALENT I-131 for the Thirteenth Operating 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.

APPLICABILITY: MODES 1, 2, 3, 4, AND 5 ACTION:

MODES 1, 2, AND 3*:

a. For the Thirteenth operating Cycle only, with the specific activity of the primary coolant greater than 0.25 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 />.

b. For subsequent cycles, with the. specific activity of the primary 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 Tavg less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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

• With Tavg greater than or equal to 500 F.

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

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REACTOR COOLANT SYSTEM ACTION: (Continued)

MODES 1, 2, 3, 4, AND 5

a. For the Thirteenth Operating Cycle only, with the specific activit'y of the primary coolant greater than 0.25 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 t within its limits.

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b. For subsequent cycles, with the specific activity of the primary  ;

coolant greater than 1.0 microcurie per gram' DOSE EQUIVALENT I-131 l or greater than 100/E microcuries per gram, perform the sampling and 4 analysis requirements of item 4a of Table 4.4-4 until the specific j activity of the primary coolant is restored to within its limits. ,

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SURVEILLANCE REQUIREMENTS

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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 3 4.4-4. '

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.a a 4s m a 7e a so too PERCD(T OF RATED THERMAL PCurER  !

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RGURE 3A 1 i DOSE EQUIVALENT I-131 Tr'.. i Coolant Specific Ac fvity Umit Versus .

7--.; of RATED THERMAL POWER with the 7ck.-i Coolant Speerne Ac:ivrty > 1.OuC1/ gram Deze Equivalent 1-131 i i

(Activity >.25/tCi/ gram Dose Eouivalent I-131 for Cycle 13 only.) ,

• 1 FARLEY-UNIT 1 3/4 4-26 AMENDMENT NO.

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REACTOR COOLANT SYSTEM j BASES i

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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 j surveillance of the conditions of the tubes in the event that there is -f 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 f measures can be taken. .l t

The plant is expected to be operated in a manner such that the secondary i coolant will be naintained 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 e.ttent 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  ;

operation will have an adequate margin of safety to withstand the loads  ;

imposed during normal operation and by postulated accidents. Operational 4 leakage of this magnitude can be readily detected by existing Farley Unit 1 j radiation monitors. Leakage in excess of this lindt will require plant '

shutdown and an unscheduled inspection, during which the leaking tubes will be  ;

located and plugged or repaired.

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For the Thirteenth Operating Cycle only, the repair limit for tubes with flaw  !

indications contained within the bounds of a tube support plate has been. l provided to the NRC in Southern Nuclear Operating Company letter dated i December 09, 1993. The repair lindt is based on the analysis contained in l WCAP-12871, Revision 2, "J. M. Farley Units 1 and 2 SG Tube Plugging Criteria l for ODSCC at Tube Support Plates," and documentation contained in EPRI Report  !

TR-100407, Revision 1, "FWR Steam Generator tube Repair Limits - Technical l Support Document for Outside Diameter Stress Corrosion Cracking at Tube ,

oupport Plates." The application of this criteria is based on limiting l primary-to-secondary leakage during a steam line break. to ensure the t applicable Part 100 limits are not exceeded.  !

. Wastage-type def ects 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. j Plugging or repair will be required for all tubes with imperfections exceeding. t 40% of the tube nominal wall thickness. If a sleeved tube is found to have j 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 204 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 B3/4 4-3 AMENDMENT NO. ..

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I REACTOR COOLANT SYSTEM $

BASES .I

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a. Mechanical l

1. Indications of degradation in the entire length of the sleeve ^ f 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. j the upper joint and the top of the lower roll expansion does  ;

not-require that the tube be removed from service.

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3. The tube plugging limit continues to apply to the portion of i 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. i

4. The tub'e plugging limit continues to apply to that portion of --

the tube above the top of the upper joint.  ;

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b. Laser Welded I

1. Indications of degradation in the length of the sleeve i between the weld joints must be evaluated against the sleeve 7 plugging limit.

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2. Indication of tube degradation of'any type including a j complete break in the tube between.the upper weld joint and [

the lower weld. joint does not lequire that the tube be '

removed from service. +

i 3 .- At the weld joint, degradation must be evaluated in both the  !

sleeve and tube. i

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

5. The tube plugging limit continues to apply to.the portion.of j

the tube above the upper weld joint and below the lower weld joint. {

Steam generator tube inspections of operating plants have demonstrated the '!

capability to reliably detect wastage type degradation that has penetrated 20%  :

of the original tube wall thickness. .!4 Whenever the results of any steam generator tubing inservice inspection fall  !

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,11f I necessary. ,

EARLEY-UNIT 1 B3/4 4-3a AMENDMENT NO. l- 1 i

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REACTOR COOLANT SYSTEM BASES

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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 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 rupture 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

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specific site parameters of the Farley site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

For the Thirteenth Operating Cycle only, the limitations on the specific activity of the primary coolant have been reduced. The reduction in specific activity. limits continues 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 snall 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 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.

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

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'i Significant Hazards Consideration Evaluation f i

in Support of the l Technical Specification Changes Associated With Steam Generator Tube Support Plate l Interim Repair Critena j

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Joseph M. Farley Nuclear Plant - Unit 1  :

Technical Specification Changes Associated With -

Steam Generator Tube Support Plate Interim Repair Criteria Sionificant Hazards Consideration Analysis INTRODUCTION  !

In letters dated January 29,1992 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 altemate repair criteria in time for use in the 1992 outages. On October 8, ,

1992, the NRC approved a 1 volt interim plugging criteria amendment for one cycle of operation of Farley Unit 1. This one cycle amendment is not valid beyond the next refueling l' outage scheduled for the spring of 1994. As a result, Southern Nuclear is proposing use of a 2 volt interim repair criteria on Farley Unit 1 for the next operating cycle. 5 DESCRIPTION OF CHANGES '

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

license amendment to implement the interim repair criteria for tube support plate elevations  !

involves no significant hazards. The interim repair criteria involves a correlation between eddy .!

current bobbin proba sigr at amplitude (voltage) and the tube burst and leakage capability.

i 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 acceptable. A sampling program would also be implemented to  !

ensure other forms of degradation are not occurring at the tube support plates and that cracks are not being masked at tube support plates by other factors.

l The proposed amendment would modify Technical Specification 3/4.4.6 " Steam Generators" '

and its associated bases, 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 l structuralintegrity. For Unit 1, the operational leakage requirement has been modified to reduce the total allowable primary-to-secondary leakage for any one steam generator from 500 l gallons per day to 140 gallons per day. In addition, the technical specification limit for specific ac'Jvity of dose equivalent l* and its transient dose equivalent l* reactor coolant specific  ;

activity is being reduced by a factor of 4 in order to increase the allowable leakage in the event j of a steam line break.

EVALUATION l

Steam Generator Tube Intearity in the development of the interim 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 1 Tubes," are used 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

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Joseph M. Farley Nuclear Plant - Unit 1 Page 2  !

Significant Hazards Consideration Analysis 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 f actors on loads for tube burst that are ,

consistent with the requirements of Section til of the ASME Code. For the tube support plate i elevation degradation 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 preserte 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 i may not function to provide a similar constraining effect during accident condition loadings. .

The WCAP-12871 analyses for Farley Unit 1 with corroded and packed crevices, as confirmed l by bobbin coilinspection, show that the tube support plates would not be significantly ,

displaced even under steam line break loading conditions. For conservatism in the interim 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 condition loadings can be satisfied with bobbin coil signal amplitudes several times larger than the proposed 2.0 volt interim 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 interim 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 interim 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 are short, tight, axially oriented micro cracks [

often 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 j 3/4 inch tube. Voltage correlation to 7/8 inch tubing size would result in an expected voltage of l 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 i 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 8.0 volts. The threshold of significant leakage (20,31/ hour or 10 '

l gpm) in a 7/8 inch tube diameter is 6 volts. -j i

- AdditionalConsiderations  !

J 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 j the steam generator by keeping structurally sound tubes in service and not unnecessarily j 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

. = . .

Joseph M. Farley Nuclear Plant - Unit 1 Page 3 Significant Hazards Consideration Analysis 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 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 reduction in a margin of safety.

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

1) Operation of Farley Unit 1 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident 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 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 normal operating conditions by the presence of the tube support plate.

Furthermore, 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 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. Based on the existing data base, this criterion is satisfied with bobbin coilindications with signal amplitudes over twice the 2.0 voit interim repair criteria, regardless of the indicated depth measurement. This structurallimit is based on a lower 95% confidence levellimit of the data. The 2.0 volt criteria provides an extremely conservative margin of safety to the structurallimit 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 with a 10 volt correlation to 7/8 inch tubing (as compared to the 2.0 volt proposed interim tube repair limit). Thus, the proposed amendment does not involve a significant increase in the ,

probability or consequences of an accident.  !

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Joseph M. Farley Nuclear Plant - Unit 1 Page 4 i Significant Hazards Consideration Analysis ,

l 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 Extemal 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 Rod 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 steam release 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 j

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 interim repair 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. Data indicate that a threshold voltage of 2.8 volts would 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 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 i 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 j factor related to the probability of detection of the flaws. If it 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 consequences of the most limiting I 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. l 1

Implementation of the proposed interim tube support plate elevation steam j generator tube repair criteria does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism 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 repair criteria has been applied (during all plant conditions). The bobbin probe signal amplitude repair criteria is l

Joseph M. Farley Nuclear Plant - Unit 1 Page 5 Significant Hazards Consideration Analysis established such that operational leakage or excessive leakage during a postulated steam line break condition is not anticipated. Southem Nuclear has previously implemented a maximum leakage rate limit of 140 gpd per steam generator on Unit 1. 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 140 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 are used below to define the " longest permissible crack" for evaluating operating leakage limits.

The single through-wall crack lengths that result in tube burst at 1.43 times steam line break pressure differential and steam line break conditions are about 0.53 inch and 0.84 inch, respectively. Normalleakage 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.

An operating leak rate of 140 gpd per steam generator has been implemented on Unit 1. This leakage limit provides for detection of 0.4 inch long cracks at nominal lea'K rates and 0.6 inch long cracks at the lower 95% confidence level leak rates. Thus, the 140 gpd limit provides for plant shutdown prior to reaching critical crack lengths for steam line break conditions at leak rates less than a lower 95% confidence level and for three times normal operating pressure differential at less than nominalleak rates.

Based on the above, the implementation of interim 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 interim tube support plate elevation repair criteria is demonstrated to maintain steam generator tube integrity commensurate with

Joseph M. Farley Nuclear Plant - Unit 1 Page 6 Significant Hazards Consideration Analysis 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 i 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 interim repair criteria is comparable ,

to that currently provided by existing technical specifications. (

in addressing the combined effects of LOCA + SSE on the steam generator 4 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 lateralloads at the wedge supports at the periphery of the plate due to either the LOCA i 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 r collapse of steam generator tubing reduces the RCS flow area through the I 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 through-wall  ;

cracks in tubes could progress to through-wall cracks during tube deformation ' j or collapse or that short through-wall indications would leak at significantly higher leak rates than included in the leak rate assessments. ,

1 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 reactor coolant system primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently low that they l need not be considered in the structural design basis of the plant. Excluding j breaks in the RCS primary loops, the LOCA loads from the large branen line - >

'! i

=

, v- .,

Joseph M. Farley Nuclear Plant- Unit 1 Page 7  !

Significant Hazards Consideration Analysis  !

breaks were analyzed at Farley Unit 1 and were found to be of insufficient magnitude to result in steam generator tube collapse or significant {

deformation. j i

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 j 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 even' ~ not significant. In actuality, the amount of secondary-to-primary leakage in me event of a LOCA + SSE is expected to be less than that currently allowed, i.e.,500 gpd per steam gererator. Furthermore, secondary- l 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 t educe 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 elevations having outside diameter stress corrosion cracking indications, l reduced operating leak rate limits, eddy current inspection guidelines to ,

provide consistency in voltage normalizMion, and rotating pancake coil  !

inspection requirements for the larger indications left in service to characterize ,

the principal 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 which must be taken out of service l 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 repair criteria 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 plant safety as defined in the .

Final Safety Analysis Report or any bases of the plant Technical  ;

Specifications.  !

CONCLUSION I

Based on the preceding analysis, it is concluded that using the interim 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 Hazards Consideration as defined in 10 CFR 50.92.

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