ML20086B941
| ML20086B941 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/27/1995 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-95-03, GL-95-3, ULNRC-3226, NUDOCS 9507060193 | |
| Download: ML20086B941 (9) | |
Text
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1901 Chouteau Avenue Post Omce Box l49 St Iouis, Missouri 63166 314-554-2650
\\
YT DonaldF.Schnen sJNION senior vice eresisent i
.Etscraic nucie.,
h3 June 27, 1995 i
U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk i
Mail Station Pl-137 l
Washington, D.C.
20555 Gentlemca:
ULNRC-3226 1
CALLAWAY PLANT DOCKET NUMBER 50-483 l
CIRCUMFERENTIAL CRACKING OF STEAM GENERATOR TUBES
Reference:
NRC GENERIC LETTER 95-03 dated April 28, 1995 The attachment to this letter contains Union Electric's response to NRC Generic Letter 95-03.
If l
you have any questions concerning this information, please contact us.
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l Very truly yours, n
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(
onald F. Schnell i
TWP/pir Attachment (6 pages) 1 J
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9507060193 950627 PDR ADOCK 05000483 p
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I STATE OF MISSOURI
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Donald F. Schnell, of lawful age, being first duly sworn upon oath says that he is Senior Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed'the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
By e
Donald P. Schnell Senior Vice President Nuclear SUBSCRIBED and sworn to before me this c2 day of
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1995.
//
M / Add.
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P' f/ f/
BARBARA J. PFAFli NOTARY PUBUC-STATE OF MISSOURI MY COMMISSION EXPlRES APRIL 22,1997 ST, LOUIS COUNTY.
l
. cc:'
T. A. Baxter, Esq.
Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.
Washington, D.C.
20037 M. H. Fletcher Professional Nuclear Consulting, Inc.
19041 Raines Dr Derwood, MD 20855-2432 l
M. J.
Farber Chief, Reactor ProjectsSection III A U.S. Nuclear Regulatory Commission i
Region III i
801 Warrenville Road Lisle, IL 60532-4351 Bruce Bartlett Callaway Resident' Office i
U.S. Regulatory Commission RR#1 Steedman, MO 65077 L. R. Wharton (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission l
1 White Flint, North, Mail Stop 13E21 l
11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department Missouri Public Service Commission P.O.
Box 360 Jefferson City, MO 65102 4
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ULNRC-3226 Page l of 3 ATTACHMENT TO-ULNRC-3226 UNION ELECTRIC RESPONSE TO NRC CENRRIC T&PrxR 95-03 l
INTRODUCTION Recent nondestructive examination of the steam generator tubing at the Maine Yankee Nuclear Plant has identified a large number of circumferential indications in the regions at the top of the tubesheet.
These inspection findings, coupled with previously documented inspection results regarding circumferential cracking, led to the issuance of NRC Generic Letter 95-03, 'Circumferential cracking of Steam Generator Tubes".
The information detailed herein addresses the requested actions of Generic Letter 95-03 as they pertain to Westinghouse designed and manufactured steam generators installed at callaway Plant.
Steam Generator Description Callaway Plant is equipped with Westinghouse Model F steam generators, the first of this type manufactured by the supplier.
The units are tubed with 11/16 inch O.D.
Inconel 600 material.
The 10 inner (short radius) tube rows are thermally treated; all remaining rows are mill annealed.
Tubing is installed with full depth hydraulic expansion in the tube sheet and all tube support plates are stainless steel with broached quatrafoil tube passages having circular lands.
Operating History Callaway Plant achieved full power operation in December, 1984.
The plant is operated on an 18-month cycle.
In-service inspection of steam generator tubing has been consistent with EPRI guidelines.
It has been our practice to inspect two of the four steam generators during each refuel outage, employing bobbin coil examination over the full length of all tubes.
Rotating pancake coil (RPC) inspection of the hot leg tube expansion transition zone has been employed on a sample basis, generally in the sludge accumulation area.
We normally examine about 300 tubes per generator with RPC.
Until Refuel 7 in spring of 1995, wear at the anti-vibration bars was the only tube degradation mechanism identified in the callaway generators.
Tube wear in this region is an expected occurrence in new generators and as predicted, the rate of wear has diminished over the years.
Table 1 provides a summary of Callaway tube examinations and plugging for the steam generators.
Industry studies of recirculating type steam generators employing alloy 600 mill annealed tubing identified the potential for hot-leg primary side stress corrosion cracking (PWSCC) in the expansion transition zone.
Studies specific to Callaway Plant i
predicted the onset of PWSCC in one percent of the tubes within i
s ULNRC-3226 Page 2 of 3 l
approximately seven effective full power years (EFPY) of service.
As a-preventive measure, in 1992, after operating for 6 EFPY, all hot leg transitions were shot peened.
In April, 1995,-during Callaway's seventh refueling, RPC sampling inspections identified the presence of both primary and secondary side stress corrosion cracking in one of the two steam i
generators examined.
The inspection program was immediately expanded to cover expansion zone transitions of all hot leg mill annealed tubing in all four generators.
A sampling of thermally treated tubing was also included-along with a small number of RPC inspections'at tube support plate intersections.
All cracks found were located.within mill annealed tubing in the transition zone.
Table 2 shows the 29 tubes that were identified with 1
indications at or near the top of the tube sheet.
Ten of the cracks were circumferential1y oriented.
The circumferential cracks were found in three of the four steam generators (6 in A, 3 in C, 1 in D). 'All were within an area defined by Rows 12-21 and Columns 47-80.
This is in the area of sludge accumulation and higher tubesheet temperatures.
Six of the indications are believed to be OD initiated, while the other four appear to be ID initiated.
SAFETY RISK ASSESSMENT Westinghouse has developed a tube burst correlation to assess burst resistance in tubing with throughwall circumferential crack indications. The burst correlation was then applied to define the limit on throughwall crack angles that satisfy the Reg. Guide 1.121 burst margin of 3 times normal operating differential pressure.
If measured RPC crack angles, after reduction for coil lead-in and lead-out effects (about 30')
for throughwall indications are less than this limit, it can be concluded that the indications satisfy burst margin guidelines.
Table 3 lists the limiting throughwall crack angle for 11/16 inch OD tubing, adjusted for lower tolerance limit material properties, with a 50% deep OD degradation existing over the remaining non-throughwall degraded section.
This table shows that a single, uniform throughwall crack of 247' with 50% deep OD degradation existing over the remaining 113" tube arc would satisfy the Reg. Guide 1.121 3AP burst requirement, and that a 283* throughwall crack with 50%, ep OD degradation existing over the remaining 77* arc would have a predicted burst pressure of 2560 psi. Based on industry-accepted detection thresholds for circumferential cracking, indications of this extent, if encountered in the field, would produce an RPC crack angle of 360*.
Additionally, it is expected that single, uniform throughwall cracks of the size represented in the table would experience primary to secondary leakage, alerting the plant operators.of the tube condition.
ULNRC-3226 Page 3 of 3 Extrapolating the results of Westinghouse analysis for 7/8 inch tubing, the corresponding crack lengths for single uniform throughwall cracks in 11/16 inch OD tubing without additional degradation in the remaining ligament would be consistent at the 3AP condition and greater than 283* at a Steam Line Break (SLB)
AP of 2560 psi.
Based on the similarity of the crack angles for the two tubing sizes with 50% degradation of the ligament, the limiting single throughwall crack angle which would support a burst capability of 2560 psi without degradation of the ligament would be approximately 310 to 320.
Westinghouse has not yet performed plant specific tube integrity evaluations for full depth hydraulically expanded tubing since the callaway cracks are the first to be experienced.
The extended operating period of Callaway Plant (approximately 8.6 EFPY) without incidence of circumferentially oriented degradation, until the most recent inspection, suggests that the conditions within hydraulically expanded tubing do not represent a potential for rapid tube degradation.
If rapid degradation were to occur, it is expected that primary to secondary leakage would be detected prior to the tube representing a burst concern at SLB recovery conditions.
Since the circumferential crack indications in callaway Alloy 600 mill annealed tubing are the first detected in full depth hydraulically expanded tubing, and considering the extended operating period of Callaway up to the time of initial crack detection, it is reasonable to conclude that crack growth rates are not a safety concern.
When considering growth rates and when factoring in the industry accepted detection thresholds for throughwall circumferential degradation, no indications would be expected at Callaway which would challenge tube integrity prior to the end of the current operating cycle.
Considering the historical performance of hydraulic expansions, it is reasonable to conclude that tube structural integrity will be maintained during all future operating cycles, provided adequate inspection programs are followed.
FUTURE TNSPECTION PLANS Callaway's next inspection will be performed during the eighth refueling outage scheduled for October / November 1996.
During this outage, 100% of the in-service hot leg expansion transitions, including the thermally treated tubing, will be examined in all four steam generators.
This examination will be performed using a qualified technique in accordance with Appendix H of the EPRI PWR Steam Generator Examination Guidelines.
In addition, 100% bobbin coil inspection will be performed in two steam generators (those not inspected in the last outage).
All data will be analyzed by Qualified Data Analysts.
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TABLE 2 CALLAWAY REPURT. 7 TUBE SHRRT INDICATIONS Steam Steam Steam Steam Generator Generator Generator Generator Indication A
B C
D Totals 4
ID Cire 4
0 0
0 4
Crack ID Axial 5
0 2
1 8
Crack Unknown 1
0 2
0 3
Axia]
OD Circ 2
0 3
1 6
Crack OD Axial 0
0 4
0 4
Crack Volumetric 0
0 5
1 6
Indication Total tubes 11*
0 15*
3 29 with indications T-Hot 616.6 613.8 617.8 615.5
- One tube in generators A and C had both a circumferential and an axial crack.
i TART,W 3 11/16 inch OD Tubing, Bnd of Cycle (BOC)
Structural Limits for Circumferential1y Oriented Degradation Single Throughwall
[
Cire. Crack with 50%
OD Degraded Licament 3AP =.3750 psi 247 Steam Line Break l
(SLB) AP = 2560 psi 283 i
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