ML20086A512
| ML20086A512 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/26/1995 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-95-03, GL-95-3, NUDOCS 9507030265 | |
| Download: ML20086A512 (14) | |
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.p Southom Nuclear Operating Company :-
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~ Post Omco Box 1295 4
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Southern Nudear Operating Company i
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Vice President
- Farley Project the southem electnc System :
l June 26, 1995 j
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Docket No.
50-348
- i 50-364 j
U. S. Nuclear Regulatory Commission j
ATTN.: Document Control Desk i
_Washipston, D.C. 20555 Joseph M. Farley Nuclear Plant
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Pa ana=a to Generic Letter 95-03 i
Ladies and Gentlemen-l On April 28,1995, the NRC issued Genenc Letter 95-03, Circumferential Crackmg in Steam
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Cc.-.;cr Tubes. The Generic Letter requested several specific actions oflicensees to address the i
threat of circ 1.waial cracking in steam generator tubes.
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Genenc letter 95-03 requested that all addresseesi
!l (1)
Evaluate recent operatmg expenence with respect to the detection and sizing of l
circumferential indications to detennine the applicability to their plant.
,1 (2)
. On the basis of the evahmhan in item (1) above, past inspection scope and results, 1
susceptibility to circumferential crackmg, threshold ofhiaa, expected or inferred crack l
growth rates, and other relevant factors, develop a safety assessment justifying continued i
operation until the next scheduled steam generator tube inspections are performed.
(3)
Develop plans for the next steam generator tube inspections as they penain to the detection of circumferential cracking The inspection plans should address, but not be limited to, l
scope (including sample expansion criteria, if applicable), methods, equipment, and criteria i
(including periu..cl traming and qualification).
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Based on these requested actions, all addressees are requested to submit:
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(A) a safety assessment justifying continued operation that is based on the evaluations j
performed in acwid.ce with Requested Actions (1) and (2) above.
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(B) a summary of the ia==~*iaa plans developed in accordance with Requested Action (3) -
1 above and a schedule for the next planned inspection.
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l 300G73 9507030265 950626 PDR ADDCK 05000348 i
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U. S. Nuclear Regulatory Comnussion
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j The safety assessment required by (A) is provided in Attachnunt 1. The summary ofinspection plans for Unit I and 2 required by (B) is provided in Attachment 2.
1 If there are any questions, please advise.
f Respectfully submitted,
.i SOUTHERN NUCLEAR OPERATINO COMPANY j
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Dave Morey REM /maf:9503REMB. DOC Attachments SWORN TO AND SUBSCRIBED BEFORE ME cc:
Mr. S. D. Ebneter THIS dd DAY
'121995 EIE.*d%s' 7>dd M o M 9 d Mr. T. A. Reed
%tary Public
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My Commission Expires:
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Safety Assessment fr 9
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4 Safety Assessment Outline 1.0 Intrathetuw 1.1 Histoncal Circumferential Degradaten Locatens 1.1.a Circumferential Degradation at the Tube Sheet Expansion
.1.1.b Circumferential Degradation at Small Radius U-Bends 1.1.c Circumferential Degrad* tion at Tube Su.pport Plates 1.1.d Circumferential Degradation of Laser Welded Sleeves 2.0 Safdy Assessment i
2.1 Tube Sheet Expansions l
2.1.a Structural Evaluation for Circumferential Cracking at the Tube Sheet Expansions i
F 2.1.b Inspections at the Tube Sheet Expansions 2.1.c Tube Integrity Assessments for Circumferential Cracking at the Tube Sheet l
Expansions l
2.2 Small Radius U-bends l
2.2.a Inspections at the Small Radius U-bends I
2.3 Dented tube support plate Intersections 2.3.a Inspections at Dented tube support plate Intersections 2.4 Laser Welded Sleeves 2.4.a inspections of Laser Welded Sleeves 2.5 Safety Assessment Sununary j
3.0 Defense in Depth: Farley Nuclear Plant Steam Generators j
4.0 Summary i
Safety Assessment l
1.0 INTRODUCTION
On April 28,1995, the NRC issued Generic letter 95-03, Circumferential Cracking in Steam Generator i
Tubes. 'lhis letter was primarily written in response to eddy current i==,W results from the Maine Yankee Nuclear Plant during the Spring 1995 outage. Maine Yankee uses Combustion Engmeering steam generators with tubes installed by the EXPLANSION process In companson, Farley Unit I uses WM=y-wse Series 51 steam generators with an explosive (WEXTEX) tube expansion. Farley Unit 2 uses Westinghouse Series 5I steam generators with full depth hardroll expansion. Both Farley units use Alloy 600 mill annealed (MA) tubing with dunensions of 0.875" OD x '
i 0.050" nominal wall thickness
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t Successive rotatmg pancake coil (RPC) probe inspection results for Westinghouse plants with hardrolled or explosively expanded tubes have indicated steadily declining numbers ofindications, declining angular l
extent, and very low growth rates. The only occurrences of significant levels of circumferential cracking have been found in plants performing the first large scale RPC inspection.
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1.1 Historical Circumferential Degradation Locations i
i Available historical infonnation shows that for some Westinghouse plants, circumferential corrosion has j
been detected in the tubesheet region at tube expansion transitions from expanded to unexpanded tubes; at the small radius U-bends; and at dented tube suppoit plate intersections (at North Anna only).
Circumferential cracking has not been detected in laser welded sleeves, the only type of sleevmg installed at Farley Nuclear Plant.
1.1.a Circumferential Degradation at the Tube Sheet Expansion
' At both Farley units, all hot leg tubes have been inspected at the top of the tubesheet region since 1990 using the RPC probe. The 100% inspection of all hot leg expansions will continue through at least the next inspections on Unit I and Unit 2. Currently available probes, coupled with properly implemented analysis i
criteria and techniques have been demonstrated to be sufficient to identify circumferential indications in the -
tubesheet region.
i 1.1.b Circumferential Degradation at Small Radius U-bends The incidence of circumferential indications at the small radius U-bends has not been significant throughout j
the industry, both in numbers ofindications and indicated arAeths, due partially to a preventive plugging i
program. Farley Unit I and Unit 2 small radius U-bend tubes were preventively plugged in the early
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1980s. Since recovery (return to senice) of the preventively plugged Row I U-bends in Farley 2 in 1990 (2R7) and in Farley 1 in 1991 (1 RIO), a 100% inspection program for exanunation of the small radius U-bends with U-bend RPC probes has been in effect up to the Farley Unit 21995 (2R10) outage. U-bend heat treatment was applied at Farley Unit 1 in 1991 and at Farley Unit 2 in 1990 in order to minimize the susceptibility of the U-bends to cracking. No circumferential cracking has been detected in the small radius U-bends since their return to senice following heat treatment.
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Safety Assessment Page 2 -
1.1.c Circumferential Degradation at Tube Support Plates l
De North Anna units which experienced circumferential indications at dented tube support plate intersections have since had their steam generators replaced. To prevent the masking of crack-like :
l indiarians in dented tube support plate intersections at Farley, all bobbin indications with dent signals I
greater than 5 volts have been RPC b = H since the Spring of 1992. No circumferential flaws have been detected at the tube support plates in the Farley steam generators.
On an induarry level, a leakage event occurred in 1987 which resulted in a steam generator tube rupture due.
l to high cycle fatigue at a desited tube at the top support plate. All domestic Westmghouse steam generators l
with carbon steel tube support plates have been analyzed for the potential of experiencing high cycle fatigue 1
at this lamtma using a methodology accepted by the NRC. In cases where the analysis indicated that
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fatigue usage could exceed 1.0, the tube was either stabilized and plugged or plugged using a leak limiting i
sentinel plug. Bree conditions must be present for high cycle fatigue at the top tube support plate;.
j dentmg, lack of antivibration bar (AVB) support, and locally elevated steam velocities due to nonuniform i
AVB inseman depths By letter dated October 30,1989, the NRC issued a safety evaluation which 1
concluded actions taken at Farley resolved the issues associated with this event and were acceptable.
-J 1.1.d Circumferential Degradation of Laser Welded Sleeves Circumferential cracking has not been detected in laser welded sleeves, the only type of sleevmg installed at I
Farley Nuclear Plant. He first exammation of the sleeves was conducted with crosswound probes at Farley 2
.i in 1993; no degradation was observed The first Farley I sleeve inspection was performed in 1994, using crosswound probes; once agam no indications of degradation were observed. After qualification (PWR Steam Generator Examination Guidelines: Revision 3, Appendix H) of the Cecco-5 probe for sleeve exammation in 1994, all laser welded sleeves in steam generator A were inspected in March 1995 at Farley 2; no indications of parent tube or sleeve degradation due to corrosion were observed.
2.0 SAFETY ASSESSMENT.
2.1 Tube Sheet Expansions The WEXTEX tube expansion process utilized in the Unit I steam generators uses an explosive charge to produce tube to tubesheet contact throughout the tubesheet region. He WEXTEX process has been implemented only in Alloy 600 MA tubing. All but one of the circumferential indications found at Farley Nuclear Plant have been in Unit 1 (WEXTEX). His structural integrity discussion is also directly applicable to the hardroll expansion in Farley Unit 2 steam generators. Plant specific assessments for both units are given in j
this section.
2.1.a Structural Evaluation for Circumferential Crackmg at the Tube Sheet Expansions l
To permit a rapid scoping assessment for tube burst capability of circumferential indications, a burst I
correlation was developed by Westinghouse for throughw311 circumferential indications. The burst correlation was then applied to define the structural limit on throughwall crack circumferential extent that satisfies the Regulatory Guide (RG) 1.121 burst margin for 3 times normal operating differential pressure.
i If measured RPC are lengths, after reduction for coil lead-in and lead-out afTects (about 30*) for
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throughwall subcations are less than the structural limit, it can be readily concluded that the indications satisfy burst margin guidelmes. If the measured arc lengths exceed the assumed throughwall structural--
limit, additional aspection (such as UT) or structural analysis are needed to assess structural integrity.
His section describes the dW.a.: of the throughwall arc length structural limit.
l 2 A development program was maartad by the WOG during the time period (1987-1992) when -
cir-r- ' ' =- ny onented degradation began to appear in WEXTEX expanded tubes. In this program, the crack simulation was psiiviw.sd by slitting tube samples using an electro discharge machined (EDM) >
process A sealing bladder with thin reinforcing foil was used to prevent premature bladder extrusion thmugh the EDM slit. EDM crack simulations and subsequent burst testing were performed for single throc3 wall cracks, segn= dad crack networks, and complex crack networks. The single EDM slit test h
results would be considered to be a conservative representation of the actual PWSCC morphology of ~
WEXTEX expansion transition cracks since the available tube pull results indicate limited throughwall crack extait (typically much less than the total measured arc length) and a more :-ge=t-d crack network
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with multiple initiation sites around the circumference. The nondegraded ligament sections betw.
corroded areas would add significantly to the burst pressure capability of an actual tube.
In the steam generator, lateral support provided by the first tube support plate restrains bendmg of the tube during pressurization and significantly adds to the burst pressure capability. Additional burst pressure capability would be provided if the tube is axially constrained at the tube support plate due to corrosion product buildup for plants with drilled hole carbon steel tube support plates. The tube support plate intersections at Farley Nuclear Plant are considered packed by tube support plate corrosion products, and as such the listed end of cycle values would be conservative for Farley Nuclear Plant, based on the added burst strength component due to axial constramt at the tube support plates. The burst pressure correlations were developed based on burst tests with lateral but not axial restraint. This data is applicable to 7/8 inch tubes a==W using a mechanical rolling process or hydraulic expansion process. Based on comparisons of burst tests for = g;=c d cracks, an analytical deternunation of the effect of nondegraded e
ligaments on the burst pressure capability of circumferential cracks was performed to develop a burst correlation for segmented cracks (with ligaments).
Utilizing the burst correlations developed from EDM data and analytical models, the' structural limits for throughwall circumferential indications were developed as given for the crack models in the following table.
The burst pressure data were adjusted to account for lower tolerance limit material properties.
7/8 Inch Tubing End Of Cycle Structural Limits for Circumferentially Oriented Degradation Single Throughwall Single'IW Crack with Segmented Throughwall Crack Model 50% Degraded Ligament Crack Model 3AP = 4500 psi 210*
210*
264*
3AP = 4300 psi 226*
226*
269*
SLBAP = 2560 psi 321*
283*
318' l
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Safety Assessment Page 4 -
' De single throughwall crack model is applicable to both ID and OD degradation. %e c,...c..:cd model is more j
typical of PWSCC. De throughwall plus 50% deep model was developed to represent 360* indications i
frequently found for ODSCC.
h is infonnative to examme the burst data of North Anna Unit I tube R18 C36. Measured burst pressure was 9250 psi and macrocrack arc length was 128* with 53' of throughwall corrosion and an average macrocrack E
depth of 90% throughwall. W +N individual data points of the EDM burst data, it is seen that a single i
throughwall EDM slit of 116* burst at 7500 psi while a -FMerack of 155' total crack arc length l
- (includung ligaments) burst at 9500 psi. %us, it is seen that the segmented EDM burst data is more j
repe==sative of actual PWSCC networks in WEXTEX transitions than the single, uniform throughwall crack
.i model,'and that use of the single, uniform throughwall crack model is very conservative.
2.1.b Inspections at the Tube Sheet Expansions
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Farley Nuclear Plant was among the first plants to perform extensive examinations of the expanded region of the steam generator tubes. Early versions of pancake array probes, specifically 8xl probes were used for the tubesheet region of Farley 1 in 1985. No indications of degradation were reported at that time, but it was regarded that the signal to noise characteristics were of such an order that little improvement over bobbin probes I
was realized. While rotating pancake probes were evolving during the late 1980's into practical inspectxm l
devices, application on a limited basis was made to characterize tubesheet signals obtamed from bobbin data.
When it became practical in 1990 to use RPC probes for large inspection programs,100% of the hot leg Farley l
2 tubesheet transitions (hardroll) were inspected A large number of axial indications were identified, and i
approxunately 325 tubes were plugged, all for axial PWSCC. During the ensuing inspection of the Farley I steam generators in 1991,100% examination of the hot leg WEXTEX transitions with RPC probes resulted in detection of 36 circumferential PWSCC indications. All tubes with arc lengths greater than 130*,1I in number, i
were stabilized prior to plugging. The largest arc length observed was 322*; this tube was examined by i
ultrasonic testing (UT) to determine whether the observed indication was continuous or had sufficient ligaments to meet tube structural integrity requirements. The UT results established the presence of a greater than 50% -
ligament in the center of the crack and additional undegraded ligaments within the crack network of about 18*.
j Tube integrity assessments performed indicate that the expected burst pressure would satisfy the RG 1.121 structuralintegrity recc....cedations.
During the subsequent refueling outages in both plants - 1992 for both units,1993 and 1995 for Unit 2, and j
1994 for Unit 1, the practice of performing 100% RPC examinations of the hot leg expansion transitions was continued. While additional indications in lower numbers were observed in each successive inspection, l
comparison of the later RPC data with data from earlier inspections demonstrated that very few new indications were being observed each year. Moreover, a low rate of PWSCC circumferential progression was evidenced by the small amplitudes observed for most of the indications.
The maximum arc lengths of the circumferential indications observed in recent' outages were 181* in Unit I i
(1RI1, October 1992) and 122'in Unit 2 (2R9, October 1993). By comparison with the prior inspection RPC data analyzed with the knowledge that a circumferential crack had been called in the current inspection, it was
' dA...i.~4 that these indications had grown approximately 10* in the previous cycle. In the most recent la==% February 1994 for Faricy 1 (1R12),2 WEXTEX PWSCC indications measuring 63
- and 85* were observed. At Faricy 2 in March 1995 (2R10), no circumferential indications were reported.
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Westmghouse has been the prime eddy current vendor at Farley Nuclear Plant for the outages discussed.
Efficacy of the probes and analyst evaluation criteria utilized can bejudged against pulled tube data from l
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another Westmghouse plant (1991) where similar analyst evaluation criteria and RPC probes as those used at Faricy Nuclear Plant were used. The RPC probe performance was judged against tube pull destructive j
exammation results. Based on this comparison, capability for detection of circumferential degradation of 23* for i
100% TW cracks and 50* (minimum) for 50% TW cracks was verified.
2.1.c Tube Integrity Assessments for Circumferential Crackmg at the Tube Sheet Expansions j
i As noted, the WEXTEX iapina guidelines were implemented in 1991. The largest circumferential l
indications were found in the first inspection implementing the WEXTEX guidelines. Four indications were found at Farley Unit I during the March 1991 inspections with RPC arc lengths exceedmg the structural limit for assumed throughwall cracks. These indications, which were 322*,271*,255*, and 245*, were UT inspected to obtam additional crack characterizatioc to support structural integrity assessments. For example, for the 322*, indication Ur identified large ligaments of about 30* and 18' with intermittent cracking (segmented crack,
.j small ligaments) of 50 to 60% depth over about 110*. Based on the large ligaments separatmg the larger i
indication into two distinct macrocracks and the large, about 50% deep wall thickness ligament, it was j
determined that this indication would be expected to provide a burst capability exceeding the structural integrity l
requirements. The other three large indications UT inspected were also found to satisfy burst requirements. For j
the last WEXTEX RPC inspection at Farley, the RPC defect arc lengths did not exceed 100".
Overall, it is concluded that all circumferential indications found in WEXTEX expansions have burst capability i
exceeding RG 1.121 burst recowencedations. The largest indications were found in the initial RPC inspections in 1991 and the number and size ofindications since 1991 has steadily decreased. This is consistent with the modest growth rates in arc lengths found for WEXTEX indications. The small arc lengths found since 1991; the decreasing trend in maximum are lengths since 1991; and operating experience showing no WEXTEX l
region primary to secondary leakage, strongly support adequate detectibility with the RPC inspections. Iflarge indications had been missed in the inspections, large arc lengths would have been found since 1991.
l Concerning growth rates, data collected at Sequoyah, another Westinghouse plant similar to Farley, suggests that WEXTEX region PWSCC does not grow at a rapid rate, and that the inspection practices sufficiently identify partial depth cracking in its early stages. No primary to secondary leakage attributed to WEXTEX l
indications was evidenced at any WEXTEX unit during at least the last two operating cycles.
As indications of circumferential degradation are plugged upon discovery, a quantified cycle to cycle growth is difficult to establish with a high degree of accuracy. A growth rate for WEXTEX indications has been
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established by examining indications having RPC inspections at the cycle detected and the prior cycle. As calling criteria are continually tightened, the chances of detecting a low level indication from a previous inspection becomes more remote. For the few numbers ofindications at Farley Nuclear Plant traceable to a i
previous inspection. circumferential crack growth rates have been estimated to be on the order of 10*.
These relatively low growth rates would not result in challenges to structural integrity between inspections and are consistent with the modest RPC crack sizes found following the initial RPC inspections.
Since no indications exceeding the structural limits have been found to date at WEXTEX expansions, the growth rates are modest, and the WEXTEX plants have not significantly increased the operating cycle lengths or operating temperatures over the prior cycle, the WEXTEX plants can continue to operate to the next scheduled inspection with no discemible risk of exceeding RG 1.121 burst capability recommendations at the end of the current operating cycles. It is expected that future operating cycles will also result in detection of circumferential cracking prior to exceeding Regulatory Guide 1.121 limits.
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j Safety Assessment Page 6 j
i 2.2 Saudi Radius U-bends Circumferential craclung has not been detected at Farley Nuclear Plant in small radius U-bends since l
preventively plugged tubes were returned to service following U-bend heat treatment. Consequently, structural cvainariana, tube integrity assessments, or growth rate analysis of U-bend circumferential cracks have not been i
> formed for Farley Nuclear Plant.
l 2.2.a inspections at the Small Radius U-bends i
Since recovery (retum to service) of the preventively plugged Row I U-bends in Farley 2 in 1990 (2R7) and in l
Farley 1 in 1991 (1RIO), a 100% inspection program for exanunation of the small radius U-bends with U-bend RPC probes has been in effect up to the Farley Unit 21995 (2R10) outage. A 100% inspection program i
continues for Unit I and the Unit 2 program is now 100% of the small radius U-bends in one steam generator only. The recovered U-bends were subjected to U-bend heat treatment to provide stress relief throughout the small radius bends; 5 of the tubes to be recovered were observed to have circumferential indications in the U-bend and were replugged. During the 1R12 inspection 2 tubes were plugged for axial PWSCC indications; no circumferential indications were observed. During the 2R10 Farley 2 examination of the U-bends in Rows !
l and 2 (steam generator A), there were no U-bend indications observed.
2.3 Dented Tube Support Plate latersections Circumferential cracking has not been detected at Farley Nuclear Plant in dented tube support plate intersections since RPC inspection of all dents with bobbin voltages greater than 5 volts was started in the Spring of 1992.
l Consequently, structural evaluations, tube integrity assessments, or growth rate analysis of circumferential cracks at dented support plates have not been performed for Farley Nuclear Plant.
l 2.3.a inspections At Dented Tube Support Plate Intersections In conjunction with the Interim Plugging Criteria (IPC), an augmented examination of tube support plate l
intersections using RPC probes was required; one of the elements of this program was intended to provide j
assurance that significant ODSCC indications were not masked by the presence of denting at some of the tube i
support phtet Since the Spring of 1002, e!! the tube 2 ppo:t plate uents > 5 volts peak to peak bobbin voltage amplitude were exanuned at both Farley units with RPC probes; no circumferential cracking was observed.
l 2.4 Laser Welded Sleeves i
Circumferential cracking has not been detected at Farley Nuclear Plant in laser welded sleeves. Consequently, structural evaluations, tube integrity assessments, or growth rate analysis of circumferential cracks in laser welded sleeves have not been performed for Farley Nuclear Plant.
2.4.a inspection of Laser Welded Sleeves Beginning with the 1992 refueling outages at Farley I and 2, many of the tube repairs required as a result of NDE examinations of the steam generator tubing have been accomplished by the use oflaser welded sleeves.
These Alloy 690 thermally treated (TF) sleeves have been installed at tube support plate elevations and in the tubesheet region. All free span laser welded sleevejoints (both attachments of the tube support plate sleeves and
Safety Assessment Page 7 the upper attachment of the tubesheet sleeve) received a post weld stress relief. The post weld stress relief for laser weld joints is designed to achieve a minimum temperature of 1400* F on the tube OD adjacent to the weld.
He first examination of the laser welded sleeves was conducted with crosswound probes at Farley 2 in 1993; no indications of parent tube degradation were observed. He first Farley I laser welded sleeve inspection was performed in 1994, using crosswound probes; once again no indications of degruation were observed. After qualification (PWR Steam Generator Examination Guidelines: Revision 3, Appendix H) of the Cecco-5 probe for sleeve examination in 1994, all 77 laser welded sleeves in steam generator A were inspected in March 1995 at Farley 2; no indications of parent tube or sleeve degradation due to corrosion were observed.
2.5 Safety Assessment Summary I
In summary, no circumferential crack has been detected in the tube sheet expansions which challenged Regulatory Guide 1.121 structural integrity requirements. Furthermore, no circumferential cracking has been detected in the small radius U-bends following their retum to service after heat treatment; no circumferential cracking has been detected in dented tube support plate intersections; and no circumferential cracking has been detected in laser welded sleeves.
3.0 DEFENSE IN DEPTII POSITIONS: FARLEY NUCLEAR PLANT STEAM GENERATORS He following assessment points further strengthen the Farley steam generator tube integrity conclusions, for the remainder of the current operating cycle:
A. The technical specification primary to secondary leak rate limit of 140 gpd for Farley I and 150 gpd for Farley 2 has been implemented into plant operating procedures. Previous evaluation by Westinghouse has established that these leak rates provide for leak before break protection for leakage assumed from a single throughwall crack leaking at mean leak rates.
B. Plant operating procedures have been revised to incorporate guidance in the event the rate of change of the leakage rate increases above acceptable levels.
C. N-16 primary-to-secondary leakage monitors have been installed in both units to provide the capability to detect and monitor rapidly changing leak rates.
D. He steam generator hot leg tube expansion transitions at Farley 2 were shotpeened in 1987 in order to relieve residual stresses. This practice is believed to have contributed to the low incidence rate of new circumferential indications.
l E. Small radius U-bend tubes preventively plugged have been recovered due to application of U-bend heat treatment. Field data has shown that this process has effectively halted corrosion of small radius U-bends. When coupled with reduced primary to secondary leakage limits and the Southern Nuclear commitment for RPC inspection of small radius U-bends, it is unlikely that rapid, unanticipated corrosion would occur a this location. No circumferential degradation has been detected in either the Farley Unit 1 or Unit 2 small radius U-bends since the tubes have been retumed to senice following U-bend heat treatment application.
Safety Assessment Page 8 F. The extensive tube examination programs at Farley I and 2 have not shown a potential for rapid, unpredicted growth on a cycle to cycle basis at any tube locations which historically have experienced SCC. Operating conditions have not changed and therefore there are no outside contributing factors which may suggest unanticipated crack growth.
l 4.0
SUMMARY
As discussed above, no circumferential indications exceeding RG 1.121 tube burst recommendations have been found at either Farley Unit I or 2. This applies to the WEXTEX and hardroll expansion transitions, laser welded sleeve joints, dented tube support plate intersections, and small radius U-bends. Since the initial 100%
RPC inspections in 1990 (Unit 1) of the hot leg tube sheet expansions, no field calls for circumferential j
indications exceeding arc lengths of 181* have been found. He current operating cycle lengths and primary temperatures are very similar to the past operating cycles for each unit at which a maximum are length of 85*
was found.
Small arc lengths found since 1991 demonstrate that the Farley steam generator RPC inspections have adequate detectibility and no structurally challenging indications have been left in service. For the few numbers of indications at Farley Nuclear Plant traceable to a previous inspection, circumferential crack growth rates have been estimated to be on the order of 10* per cycle.
Based on the previous cycle practice of 100% RPC inspection of the tubes at the top of the tubesheet elevation, resulting in either plugging or sleeving of all detected circumferential indications, an extremely low incidence rate of new indications, low apparent growth rate and strict secondary side chemistry control, no Farley tubes are projected to challenge the structural integrity limits in the current operating cycle nor are circumferential indications expected to challenge tube integrity in future cycles. Thus, the Farley Nuclear Plant steam generators can safely operate to the next scheduled inspection.
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Inspection Programs for Next Outages P
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Inspection Programs for Next Outages Schedule. Scope. and Equipment The next scheduled inspections for Unit I and Unit 2 are Fall 1995 and Fall 1996, respectively. At this time, inspection plans for the Unit i 13RF and the Unit 211RF outages include:
Farley Unit I and Unit 2 Next Subsequent inspection Plans Inspection Region Farley Unit 1 Farley Unit 2 Small Radius U-bends 100% all steam generators 100% of one steam generator using U-bend RPC using U-bend RPC Dented TSPs All dents >5 volts using RPC All dents >5 volts using RPC Laser Welded Sleeves 100% all steam generators 100% all steam generators using CECCO 5 using CECCO 5 Expansion Transition 100% hot leg,20% cold leg 100% hot leg,20% cold leg using RPC using RPC Exoansion Plans Expansion plans, where 100% inspection is not conducted, will be based on structural integrity evaluations and the nature and number of flaws discovered. The criteria and scope of the Unit I cold leg expansion transition inspection will be based on the Westinghouse Owners Group WEXTEX Committee recommendations.
Training and Oualification Analysts will be QDA qualified per Appendix G of the PWR Steam Generator Examination Guidelines (NP-6201 Rev.3), and training will be updated to include the most recent lessons leamed from tube pulls at other plants in the industry. The guidelines for interpretation of ECT signals already in use at Farley will be reinforced by correlation of the past Farley experience with the cumulative industry experience with detection and sizing of circumferential cracks.
Methods The existing program for inspection of regions potentially susceptible to circumferential cracking, i.e., small radius U-bends, dented tube support plate (TSP) intersections, and expansion transitions, already meets the intent of the EPRI recommendations as embodied in the PWR Steam Generator Examination Guidelines (NP-6201 Rev.3).
Criteria All tubes having circumferential flaws detected with the RPC probe or having circurnferential flaws detected in the pressure boundary oflaser welded sleeves will be repaired or plugged prior to retuming the Farley steam generators to service.
The above plans are subject to change if significant changes occur in NDE technology or knowledge of inspections.