ML20114B490
| ML20114B490 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/24/1992 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML19303F046 | List: |
| References | |
| NUDOCS 9208280268 | |
| Download: ML20114B490 (21) | |
Text
s Red-Lined and Revised Technical Specification Pages for the Interim Plugging Criteria w
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4.4.6.4 Acceptance Criteria l
As used in this Specification:
a.
1.
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication dravings or specifications.
Eddy-current testing indications belov 20% of the nomiaal vall thickness, if detectable, say be considered as leperfections.
2.
Degradation means a service-inducea cracking, vastage, vear 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 vall thickness caused by degradation.
4.
% Degradation means the percentage of the tube or sleeve vall thickness af fected 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.
PlJu ging or Repair t.init means the leperftetton 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 vall thickness.
For a tube that has been sleeved with a mechanical joint sleeve, through vall penetration of greater than or equal to 31% of sleeve nominal g
vall thichess in the sleeve requires the tube to be removed I
from service by plugging.
For a tube that has been sleeved with a velded joint sleeve, through vall penetration greater than or y
equal to 37% of sleeve nominal vall thickness in the sleeve between the veld joints requires the tube to be removed from service by plugging.
l 3
4 7.
Unserviceable describes the condition of a tube or sleeve if it leats or contains a defect large enougt to affect its structural integrity in the event of an operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedvater line break as specified in 4.4.6.3.e, above.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube that has l
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 Vestinghouse report VCAP-11178 Rev. 1, or laser velded sleeving, as y
described by Vestinghouse report VCAP-12672, which is used to main-tain a tube in service or return a tube to service. This includes the renoval of plugs that were installed as a corrective or preventive seasure.
FARI.EY-UNIT 1 3/4 4-12 AMENDMENT NO.
Tii,,1, 63 D
FL119R CQaLANT 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 g nerator 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 o>eration would be limited by the limitation of steam generator tube leakage setween the primary coolant system and the secondary coolant system (primary-to-secondary leakage - 140 gallons per day per
[
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 ano by postulated 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.
l 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.1.12) 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:
a.
Mechanical 1.
Indications of degradation in the pntire 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.
FARLEY-UNIT 1 83 AMENDMENT NO.
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INSERT A At tube support plate intersections, the repair limit for the Twelfth Operating Cycle is based on maintaining steam generator tube serviceability as described below:
a.
An eddy current examination using a bobbin probe of 100% of the hot and cold leg steam generator tube support plate 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 bobbin voltage less than on equal to 1.0 vnlt will be allowed to remain in service.
c.
Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 1.0 volts will be repaired or plugged except as noted in 4.4.6.4.a.6.d below.
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 l
degradation with a bobbin voltage greater than 3.6 volts will be plugged or repaired.
INSERT B For the Twelfth Operating Cycle only, the repair limit for tubes with flaw indications contained within the baunds of a tube support plate has been provided to the NRC in Southern Nuclear Operating Company letter dated August 24, 1992.
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 Suppo.t Plates."
The application of this criteria is based on limiting primary
,a-secendary leakage during a steam line break to less than 1 gallon per minute.
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I ftEACTOR COOLART SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
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4.4.6.4 Accepttnce 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.
Dearadaticn 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 tuta 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.
% 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 to 40% of the nominal tube wall thickness.
For a tube that has been sleeved with a mechanical joint fleeve, 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.
At tube support plate intersections, the repair limit for the Twelfth Operating Cycle is based on maintaining steam generator tube serviceability as described below:
a.
An eddy current examination using a bobbin probe of 100% of the hot and cold leg steam generator tube support plate intersections will be performe(i for tubes in service.
b.
Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltage less than or equal to 1.0 volt will be allowed to remain in service, c.
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 will be repaired or plugged except as noted in 4.4.6.4.a.6.d below.
FARLEY-UNIT 1 3/4 4-12 AM:NDMENT NO.
l REACTOR (QDL&NT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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.
Mnserviceable 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 Inspectipa 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 'ieg.
For a tube that has been repaired by sleeving, the tube inspection should include the sleeved portion of
+he tube.
9.
Tube Repair refers to mechanical sleeving, as described by Westinghouse report kCAP-11178, Rev.1, or laser welded sleeving, as described by hestinghouse 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 correct 4"a ar preventive measure.
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FARLEY-UNIT 1 3/4 4-12a AMENDMENT NO.
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REACTOR COOLANT SYSTEM BASES 3/4.4.6 STEAM GENERATOM 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 sur veillance 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 woJ1d be limited by the limitation of steam generator tube leakage between the primtry coolant system and the secondary coolant splem (primary-to-secondary leakage.14L gallons per day per steam generator).
Cracks having a primary-to-secondary leakage les-than this limit during operation will have an adequate margin of safety to withstand Ae loads imposed during normal operation and by postulated accidents.
Operational leakage of this magnitude can be readily detected by existing F?rley Unit I radiation monitors.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during l
which the leaking tubes will be located and plugged or repaired.
for the Twelfth Operating Cycle only, the repair limit for tubes with flaw indications I
contained within the bounds of a tube support plate has been provided to the NRC in.
Southern Nuclear Operating Company letter dated August 24, 1992.
The repair limit is based on the analysis contained in WCAP-l?871, Revision 2, "J. M. Farley Units 1 and 2 l
SG Tube Plugging Criteria for 00500 at Tube Support Plates." The application of this l
criteria is based on limiting primary-to-secondary leakage during a steam line break to less than 1 gallon per minute.
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 i
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:
l 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.
1 FARLEY-UNIT 1 B3/4 4-3 AMENDMENT NO.
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in support of the Interim Plugging Criteria 4
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Joseph M. Farley Nuclear Plant Unit 1 Steam Generator Tube Support Plate Interim Plugging Criteria Significant Hazards Consideration Analysis INTRODUCTION 1
in a letter dated 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 Unit 1 Fall of 1992 outage.
The Staff indicated a willingness to discuss an interim plugging criteria for use in the Unit 1 Twelfth Operating Cycle.
As a result, Southern Nuclear is proposing an interim plugging criteria, developed by adding additional conservatisms to the alternate plugging criteria.
DESCRIPTION OF THE AMENDMENT REQUEST As required by 10 CFR 50.91(a)(1), an analysis is provided to demonstrate that the proposed license amendment to implement the. interim plugging criteria for tube support plate elevations for farley Unit 1 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 bobbin probe voltages less than or equal i > 1.0 volt, regardless of indicated depth, do not require remedial action if gostulated 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.
The proposed amendment would modify Technical Specification 3/4.4.6 " Steam Generators," and its associated bases.
The steam generator plugging / repair limit will be modified to use an alternate plugging limit for determining serviceability for tubes with outside diameter stress corrosion cracking at the tube support plate by a methodology that more reliably assesses structural integrity.
An amendment far the operational leakage requirement has previously been submitted to reduce the total allowable primary-to-secondary leakage for any one steam generator from 500 gallons per day to 140 gallons per day. Until issuance of this amendment, this limit has been administrative 1y implemented.
The Technical Specifications and the associated Bases will indicate that these changes are applicable only for the Unit 1 Twelfth Operating Cycle.
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Significant Hazards Consideration Analysis Page 2 EVALVATION 11 tam Generator Tube intearity Discussion in the development of the interim plugging criteria, R.G. 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes," and R.G.1.83, " Inservice inspection of PWR Steam Gene,stor 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 determining tha 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 occurring in the farley steam generators, tube burst criteria are inherently satisfied during po mal operating conditions by the pre >ence 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 whether the tube support plate would function to provide a similar constraining effect during accident condition loadings in Farley Unit 1.
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 on the existing data base, burst testing shows that the safety requirements for tube burst margins during both normal and accident condition loadings can be satisfied with bobbin coil signal amplitudes less than 6.2 volts, 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 plugging criteria, tube leakage considerations must also be addressed.
It must be determined that the cracks will nat leak excessively during all plant conditions.
For the interim tube i
plugging criteria developed for the Farley Unit 1 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 macrocracks separated by ligaments of material.
No leakage during normal operating conditions has been ebserved in the field for crack indications with signal amplitudes less than 6.2 volts in a % inch tube. Voltage correlation to % inch tubing size would result in an expected voltage of 8.4 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.
Laboratory data for pulled tubes and model boiler specimens show limited leakage for indications under 10.0 volts during a postulated SLB condition.
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i Significant Hazards Consideration Analysis Page 3 i
Additional Considerations l
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 l
in the reactor coolant flow thr0 ugh the steam generator in LOCA analyses.
The proposed amendment would avoid loss of margin in reactor coolant system flow l
and, therefore, assist in demonstrating that minimum flow rates are maintained 1
in excess of that required for operation at full power.
Reduction in the i
amount of tube plugging required can reduce the length of plant outages and I
reduce the time that the steam generator is open to the containment i
environment during an outage.
I ANALYSIS (3 FACTOR TEST) i in accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license amend
.it is analyzed using the following standards and found not to: 1) involve a significant increase in the probability or i
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 definnd 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 cs 5000 psi for-indications of outer diameter stress corrosion cracking with voltage measurements as high as 30 volts.
Burst testing performed on pulled l
tubes with up to.10 volt indications show burst pressures in excesc of j
5900 psi at room temperature. Correcting for the effects of temperature on material properties and minimum strength levels (as the burst testirig was done at room temperature), tube burst capability significantly l
exceeds the R.G.1.121 criterion requiring the maintenance of a margin of three times normal operating pressure differential or, tube burst if through-wall cracks are present.
Basn on the existing data base, this criterion is satisfied with bobbin cell indications with sigral l
amplitudes less than 6,2 vo
.s, regardless of the indicated depth measurement.
This structural limit is based on a lower 95% confidence level limit of the data. The 1.0 threshold volt 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 6.2 volts so that a unique-crack length is not defined by a burst pressure to voltage correlation.
- However,
Significant Hazards Consideration Analysis Page 4 relative to expected leakage during normal operating conditions, no field leakage has been raported from tubes with indications with a voltage level of under 6.2 volts for a % inch tube, with 8.4 volts correlation to % inch tubing (as compared to the 1.0 volt proposed interim tube plugging limit).
Relative to the expected leakage during accident condition loadings, tha accidents that are affected by primary-to-secor.dary leakage and stcam release to the en/ironment are loss of External Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Seccndary 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 Unit 1 in considering the potential for offsite doses.
The offsite dose analyses for the other events which model primary-to-seconoa y leakage arid 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 differential pressure, as is the case following a steam line break event.
This increase in differential prespire is responsible for the postulated increase in leakage and associated offsite doses following a steam line break event.
Upon 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 withir the current licensing basis for Unit 1, 1 gallon per minute during a steam line break event. Data indicate that a threshold voltage of 2.8 volts would result in through wall cracks long enough to leak at SLB conditions.
Application of the proposed plugging criteria requires that the current distribution of a number of indications versus voltage be obtained during the Unit 1 Eleventh Refeeling Outage.
The current voltage is then combined with the rate ci change in voltage measurement to establish an end of cycle voltage dira ibution and, thus, leak rate during SLB pressure differential.
If it is found that the potential SLB leakage for degraded intersections plar,aed to be left in service exceeds 1 gallon per minute, then additional tubes vill oe plugged or repaired to reduce 3LB leakage potential to 1 gallon per minute or less.
2)
The proposed license amendment does 0.ot create the possibility of a new or different kind of accident from any accident previously evaluated.
Implementation of the proposed interim tube support plate elevation steam generator tube plugging criteria does not introduce any significant changes to the plant design basis.
U<a of the criteria does not provide a mechanism which could rest.it in an accident outside of the region of the tube support plate. elevations.
Neither-c single or multiple tube-rupture event would be expected in a-steam generator in which the plugging criteria has been applied (during all plant conditions).
The 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.
Significant Hazards Consideration Analysis Page 5 SNC has implemented a maximum leakage rate limit of 140 gpd per steam generator to help preclude the potential for excessive leakage during all plant conditions.
The R.G.1.121 criterion for establishing operational leakage rate limits that require plant shutdown t.re based upon leak-before-break considerations to detect a free span crack before potential e
tube rupture.
The 140 gpd limit should provide 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. I 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 longe:t 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 6.2 volts for typical ODSCC corresponds to meeting this tube burst requirement at the lower 95% uncertainty limit on the burst-correlation.
Alternate crack morphologies can correspond to 6.2 volts so that a unique crack length is not defined by the burst pressure versus vultoge correlation.
Consequently, typical burst pressure versus through-wall crack length correlations are used below to define the
" longest permissible crack" for ev luating operating leakage limits.
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 for these crack lengths would range from 0.11 gallons per minute to 4.5 gallons per minute, respectively, while lower 95% confidence level leak rates would range from about 0.02 gallons per minute to 0.6 gallons per minute, respectively.
An operating leak rate of 140 gpd has been implemented due to the 3
detection of circumferential flaws in the expansion region. 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 gpd limit provides for plant shutdown prior to reaching r"itical crack lengths for SLB conditions at leak rates less than a lower 95% confidence level _nd-for three times normal operating pressure differential at less than nominal leak rates.
3)
The proposed license amendment does not involve a significant reducHon in margin of safety.
The use of the interim tube support plate elevation plugging criteria at Farley Unit 1 is demonstrated to maintain steam generator ti e - integrity commensurate with the requirements of R.G. 1.121.
R.G. 1.12i describes a method acceptable to the NRC staff for meeting GDC 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 shoald be removed from service.
Upon implementation of the criteria, even under the worst case conditions, the occurrence of ODSCC at the tube support plate elevations is not expected to lead tc 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
Significant Hazards Consideration Analysis Page 6 precluded by verifying that, once the criteria are applied, tha expected end of cycle distribution of crack indications at the tube support plate elevations would result in mininnl, and acceptable, primary to secondary leakage during all plant conditions and, hence, help to demonstrate radiological conditions are less than a small fraction of the 10 CfR 100 guideline.
In addressing the combined effects of LOCA + *SE on the steam generatcr compenent (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 loanings.
Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.
Additionally, the margin to burst for tuber using the interim plugging criteria is comparable to that currently provided by ex, sting Technical Specifications.
There are two issues associated with steam generator tube collapse.
First, the collapse of stea,a generator tubing reduces the rcd flow area through the tubes. The reduction in flow area increases -the rcsistance 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 that partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse.
Consequently, a detailed leak-before-break analysis was performed and it was concluded that the leak-before-break methcdology (as permitted by GDC
- 4) is applicable to the f arley Unit I 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:ine structural design basis of the plant.
Excluding breaks in the RCS primary loops, the LOCA loads from the large branch line breaks were. analyzed at farley Unit 1 and were found to be of. insufficient magnitude to result in stean, generator tube collapse or significant deformation.
Regardless of whether or not leak-before-break is applied to the primary loop piping at Farley Unit 1, any flow area. reduction is expected to be minimal (much less than 1%) and PCT margin is available to account fnr this potential effect.
Based-on recent 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 level s..For all other steam generator tubes,.the possibility of secondary-to-primary leakage in_ the event _ of a LCCA + 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 currently allowei, i.e.,140 gpd per steam generator.
Furthermore, secondary-to--
Significent Hazards Consideration Analysis Page 7 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 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 vepaired.
The installation of steam generator tube plugs would reduce the RCS flow margin, thus implementation of the interim plugging criteria will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.
Based on the above, it is concluded that the proposed 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 1
Based on the preceding analysis, it is concluded that using the TSP elevation bobbin coil probe signal amplitude interim steam generator tube plugging i
criterion for removing tubes from service or repairing tubes at Farle;, Unit 1 i s acceptable and the proposed license amendment does not involve a Significant Hazards Consideration Finding as defined in 10 CFR 50.92.
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Unit 1 Eleventh Refueling Outage Proposed Eddy Current Guidelines for use with the Interim Plugging Criteria
Guidelines for Use With The Interim Plugging Criteria This attachment contains guidelines which provide direction in applying the interim plugging criteria.
The folluwing 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 Probg 1.
Bobbin Coil Probe Specification See Section A.2.1 of Appendix A to WCAP-12871, Revision 2.
2.
Bobbin Coil Calibration Standard See Section A.2.2 and A.2.3 of Appendix A to WCAP-12871, Revision 2.
3.
Bobbin Coil Data Acouisition and Analysis See Section A.2.4, A.2.5 and A.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 velt 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 Spec 1fication See Section A.2.1 of Appendix A to WCAP-12871, Revision 2.
2.
RPC Calibration Standard See Section A.2.2 of Appendix A to WCAP-12871, Revision 2.
3.
RPC Data Acouisition and Analysis All tube support intersections with bobbin coil flaw indications registering greater than 1.0 volt shall be inspected with the RPC, See Section A.3.6 of Appendix A to WCAP-12871, Revision 2.
l Guidelines For Use With The IPC Page 2 1
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RPC Flaw Recordina Guidelines 1
For TSP intersections with a bobbin flaw indication voltage greater than 1 volt, all RPC indications of flaws shall be recorded.
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1 Reportina Reauirements 1.
Southern Nuclear will inform the NRC, prior to Cycle 12 operation, of any unexpected inspection findings relative to the assumed characteristics of the flaws at the 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 l
the NRC.
2.
The predicted SLB leakage will be reported to the Staff prior to restart 4
from the eleventh refueling outage for Unit 1.
Southern Nuclear understands and accepts that the Staff's approval of the interim repair limit is contingent on the demonstration that the predicted SLB leakage at the end-of-cycle 12 will not exceed 1 gpm.
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Joseph M. Farley Nuclear Plant - Unit 1 i
Technical Specifications Changes Associated With Steam Generator Tube Support Plate Interim Plugging Criteria Environmental Evaluation Pursuant to 10 CFR 51.22(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 Unit 1 Technical Speci#ications associated with steam generator tube support plate interim plugging cr. ceria 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 Tarley Nuclear Plant.
i Based upon this evaluation it can be concluded pursuant to 10 CFR 51.22(b) that it is not necessary to perform an environmental assessment or an environmentet impact statement.
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WCAP-13464, " Response to NRC Questions on the Farley Steam Generator Tube Alternate Plugging Criterion Presentation Materials," (Proprietary) 2.
WCAP-13464, " Response to NRC Questions on the Farley Steam Generator Tube i
Alternate Plugging Criterion Presentation Materials," (Non-Proprietary) 4-k l
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