ML040970336
| ML040970336 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/05/2004 |
| From: | Hartz L Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 04-026, FOIA/PA-2005-0210 | |
| Download: ML040970336 (6) | |
Text
Dominion Nuclear Connecticut, Inc.
Mill\\rone Powcr Station Ropc Fcrty Road Warcrford, (:I' 06385 A p r i l 5, 2004 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 h
v borninion Serial No.:
04-026 NLOS/PRW Rev. 1 Docket No.:
50-423 License No.:
NPF-49 DOMINION NUCLEAR CONNECTICUT, INC., (DNC)
MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING STEAM GENERATOR TUBE INSPECTION
SUMMARY
FOR FALL 2002 OUTAGE In letters dated October 7, 2002 and August 25, 2003, Dominion Nuclear Connecticut, Inc. (DNC) submitted to the U. S. Nuclear Regulatory Commission (NRC) the Fall 2002 refueling outage steam generator tube plugging report and the corresponding steam generator tube insewice inspection summary report for Millstone Power Station Unit 3.
In a letter dated January 7, 2004, the NRC requested additional information to complete their evaluation of these reports. Attachment 1 of this letter is the response to the request for additional information.
If you should have any questions regarding this submittal, please contact Mr. Paul R.
Willoughby at (804) 273-3572.
Very truly yours, Leslie N. Ha&
Vice President - Nuclear Engineering Attachment Commitments made in this letter: None.
Serial No.: 04-026 Docket No. : 50-423 Page 2 of 2 cc:
U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1 41 5 Mr. V. Nerses Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M.
Schneider NRC Senior Resident Inspector Millstone Power Station
ATTACHMENT 1 REQUEST FOR ADDITIONAL INFORMATION STEAM GENERATOR TUBE INSPECTION
SUMMARY
FALL 2002 OUTAGE
Serial No.: 04-026 Docket No.: 50-423 Page 1 of 3 DOMINION NUCLEAR CONNECTICUT, INC.. (DNC)
MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION OUTAGE In letters dated October 7, 2002 and August 25, 2003, Dominion Nuclear Connecticut, Inc. (DNC) submitted to the U. S. Nuclear Regulatory Commission (NRC) the Fall 2002 refueling outage steam generator tube plugging report and the corresponding steam generator tube inservice inspection summary report for Millstone Power Station Unit 3.
In a letter dated January 7, 2004, the NRC requested additional information to complete their evaluation of these reports.
Below is the response to the request for additional information:
NRC Question 1 On page two of your inservice inspection report, you indicated that the rotating pancake coil examinations in Steam Generator (SG) A resulted in 43 volumetric indications. Of these indications, 28 were determined to be manufacturing burnish marks (MBMs) and the remaining 15 were considered to be MBMs or loose part wear that could not be confirmed with bobbin coil. The last 15 indications were reported as single volumetric indications (SVI) and documented in Attachment 4 of your report. There seems to be a difference between the number of SVls listed in Attachment 4 and the number of SVls discussed in the report. A similar situation occurs for the number of SVls in SG C.
Please clarify the differences.
DNC Response The number of indications reported in the inspection report is correct for each steam generator. That is, the rotating pancake coil examinations in Steam Generator A resulted in forty-three (43) volumetric indications, of which twenty-eight (28) were determined to be manufacturing burnish marks (MBMs). The remaining fifteen (1 5) were considered to be MBMs or loose part wear that could not be confirmed with bobbin coil. In Steam Generator C there were sixteen (16) volumetric indications of which eight (8) were determined to be manufacturing burnish marks. The remaining eight (8) were considered to be MBMs or loose part wear.
The applicable paragraphs of the inspection report identified that the fifteen (15) SVls in Steam Generator A and eight (8) SVls in Steam Generator C were documented in the associated attachments. In those attachments (i.e., 4 and 7) all observed SVls were identified, including duplicative observations as a result of different instrument channel frequencies (i.e., use of channels 4 and p12). The apparent contradiction noted in the question is actually the result of the individual reporting of each measured indication.
Duplication exists as each indication is reported twice, once for each channel. Line by line review of Attachments 4 and 7, requires that each SVI be checked as to channel and location.
Serial No.: 04-026 Docket No.: 50-423 Page 2 of 3 R29 -Clog R29 -C110 The associated table below specifically identifies the row and column of each tube and the elevation of each indication, as identified on channel 4 and on channel p12.
Although there were differences in the reported size of the indication (i.e., percent through-wall degradation) dependent on the channel used, the indication was categorized using the most conservative sizing.
+0.8
+0.8 0.3 0.3 0.9 0.73 0.9 0.73 NRC Question 2 On page four of your IS1 report, you indicated that one tube in SG C was identified with an obstruction. Please describe the nature of the obstruction. Include in your response a discussion of whether the obstruction was service-induced and the extent and location of the obstruction (e.g., what was the largest size probe to be passed through the tube during this outage, and previous outages).
DNC Response One tube, Row I, Column 115, in C Steam Generator, was identified with an obstruction. That obstruction occurred at 7.29 inches above the cold leg tube end or approximately 13 inches below the top of the tubesheet. The obstruction blocked the insertion of both the 560 and 540 probes. Smaller probes, such as the 520, were not attempted, as they would not have provided adequate fill factors. Consequently, the tube was removed from service. No further inspections were performed.
Historical eddy current data review determined that the tube had been inspected with a 560 probe pre-service (i.e., 1985), 3R01, 3R02, 3R04, Mid-cycle 3R06, and 3R06.
Consequently, we conclude the obstruction was service induced.
Serial No.: 04-026 Docket No.: 50-423 Page 3 of 3 SIG A
A NRC Question 3 Description Evaluation Location Above tubesheet between column 23 and 24 Above tube support # I in bundle Small diameter wire(s)
ER-99-0094 near row 42 (cold side)
Machine curl ER-99-0094 On page four of your report, you indicate that an SVI was identified and is in contact with a non-retrievable loose part. Please discuss the nature of this part and discuss the results of evaluations performed to ensure the part would not damage other adjacent tubes. In addition, please discuss what other loose parts have been left in the SGs and the results of any evaluations performed to ensure these parts would not result in a loss of tube integrity for the period of time between inspections.
A TTS+4HL, between R9-Cl07 and RIO-C107 A
R48-C48 B
Annulus of S/G tubesheet A
R58-C68 B
Above tubesheet near R74-C97 C
C Above tubesheet near R58-C47 Above tube support #2 visible from flow slot TTS at RIC4 cold lea lodaed between DNC Response CR-02-10007 CR-02-10007 C shaped machine curl Machine curl Sludge rock M3-EV-02-0035 Small metal shaving CR 002998 Sludge rock (1/8"x1/8"~1/16")
ER-99-0094 Small diameter wire ER-99-0094 ER-99-0094 Metal shaving (1 / I 6"x3/4')
Stabilized The subject foreign material, found on the secondary side of the 'C' steam generator, was categorized as a small section of flat stock which had become wedged between tubes. Review of historical ECT data identified that this small part had been at this location since 1989. Although only minor damage has been identified, conservative engineering decision-making dictated that the tube in contact with the foreign material be removed from service and stabilized. The stabilizer increases the damping of the tube, reducing flow induced vibrations, and prohibits interaction with adjacent tubes in the unlikely event of a complete severance.
C I
blowdown pice andtube The attached table identifies the current evaluation references for all known foreign material remaining in the steam generators. These references are maintained onsite for NRC review. Based on the evaluations performed, tubes were removed from service unless the evaluation determined that potential foreign material interaction would not result in unacceptable tube damage or catastrophic tube failure. Additionally, Millstone Unit 3 inspects every tube within five effective full power years in accordance with the standards established in NEI 97-06.
Consequently, the tubes are monitored and removed from service prior to developing wall loss or damage that could exceed the tube structural integrity performance criteria.
Plate CR-02-09738 D
I
\\
Possible piece of slag CR-01-01537 Cold leg distribution baffle vicinity of tubes R57-C74. R57-C75. R58-C74. R58-C75