ML17264A303
| ML17264A303 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/19/1995 |
| From: | Andrea Johnson NRC (Affiliation Not Assigned) |
| To: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| GL-95-03, GL-95-3, TAC-M92244, NUDOCS 9512220186 | |
| Download: ML17264A303 (7) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 19, 1995 Dr. Robert C. Hecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
R.
E.
GINNA NUCLEAR POWER PLANT REQUEST FOR ADDITIONAL INFORHATION (RAI) CIRCUHFERENTIAL CRACKING OF STEAM GENERATOR TUBES (TAC NO. H92244)
Dear Dr. Hecredy:
The NRC staff has reviewed your Hay 5,
- 1995, response to Generic Letter 95-03, "Circumferential Cracking of Steam Generator Tubes," for your plant, and has identified several areas where additional information is needed.
Enclosed is an RAI for R.
E. Ginna Nuclear Power Plant.
Please provide your response within 30 days from the r eceipt of this RAI.
This request is within the original reporting burden for information collection of 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> covered by the Office of Hanagement and Budget clearance number 3150-0011, which expires July 31, 1997.
Sincerely, Docket No. 50-244
Enclosure:
Request for Additional Information cc w/encl:
See next page
'l en R. Johnson, Pr 'ect Directora Divi of Reac Office o r
roject Hanager e I-l r Projects I/II Reactor Regulation 9512220186 951219 PDR ADOCK 05000244 P
Dr. Robert C. Mecredy R.E.
Ginna Nuclear Power Plant CC:
Peter D. Drysdale, Senior Resident Inspector R.E. Ginna Plant U.S. Nuclear Regulatory Cooeission 1503 Lake Road
- Ontario, NY 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Coomission 475 Allendale Road King of Prussia, PA 19406 Mr. F. Williams Valentino, President New York State
- Energy, Research, and Development Authority 2 Rockefeller Plaza
- Albany, NY 12223-1253 Charlie Donaldson, Esq.
Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Nicholas S.
Reynolds Winston
& Strawn 1400 L St.
N.W.
Washington, DC 20005-3502 Ms. Thelma Wideman
- Director, Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31
- Lyons, NY 14489 Ms. Mary Louise Meisenzahl Administr ator, Monroe County Office of Emergency Preparedness ill West Fall Road, Room ll Rochester, NY 14620
R.
E.
GINNA NUCLEAR POWER PLANT RE(VEST FOR ADDITIONAL INFORNTION The following areas have been identified as being susceptible to circumferential cracking:
'a ~
b.
c ~
d.
Please clarify what rows of tubes are considered "short radius" tube U-bends.
Expansion transition circumferential cracking Small radius U-bend circumferential cracking Dented location (including dented TSP) circumferential cracking Sleeve joint circumferential cracking In your response, areas b, c, and d were not specifically addressed although it was indicated that inspections have been performed to determine the susceptibility to circumferential cracking.
Please provide the requested information for these areas (and any other areas susceptible to circumferential cracking) per Generic Letter (GL) 95-03.
Please specify the probe used during these examinations and the extent of the examinations (e.g.,
100X of all dents with bobbin voltages greater than 5.0 volts).
The staff realizes that some of these areas may not have been addressed since they may not be applicable to your plant; however, this should be clearly stated with the basis for the statement (e.g.,
no sleeves are installed; therefore, the plant is not susceptible to sleeve joint circumferential cracking).
In your response, it was indicated that tube support plate regions were inspected.
Please clarify whether the locations sampled were dented tube support plate intersections.
Describe the criteria for determining which locations were inspected.
If a dent voltage threshold was used for such a determination, the calibration procedure used (i.e., 2.75 volts peak-to-peak on 4-20X through-wall ASIDE holes at 400/100 mix) should be provided.
In your response, it was indicated that no depth sizing has been performed for the circumferential indications detected at the expansion transition.
In addition, it was stated that no through-wall indications have been detected and that if a through-wall indication is detected it will be detected through chemistry and radiation monitoring detectors.
Please discuss whether the largest indication detected during the previous outage was capable of withstanding the pressure loadings specified in Regulatory Guide 1. 121 and the basis for this conclusion (e.g.,
tube pulls, in-situ pressure testing, etc.).
The staff notes that (I) structurally significant indications do not necessarily have to be through-wall and (2) leak-before-break is generally not relied upon for stress corrosion cracking.
Please clarify the baseline inspections to be performed on the replacement steam generators prior to plant start up.
Enclosure
6.
As a result of discovering circumferentially oriented degradation at the top of the tubesheet, other plants with partial depth roll expansions perform inspections with techniques capable of detecting circumferentially oriented degradation in this region.
If this area is susceptible to circumferential cracking, please provide the information requested in GL 95-03 (e.g.,
past inspection scope and results).
7.
During the Maine Yankee outage in July/August
- 1994, several weaknesses were identified in their eddy current program as detailed in NRC Information Notice 94-88, "Inservice Inspection Deficiencies Result ih Severely Degraded Steam Generator Tubes."
In Information Notice 94-88, the staff observed that several circumferential indications could be traced back to earlier inspections when the data was reanalyzed using terrain plots.
These terrain plots had not been generated as part of the original field analysis for these tubes.
For the rotating pancake coil (RPC) examinations performed at your plant at locations susceptible to circumferential cracking during the previous inspection (i.e., previous inspection per your GL 95-03 response),
discuss the extent to which terrain plots were used to analyze the eddy current data.
If terrain plots were not routinely used at locations susceptible to circumferential
- cracking, discuss whether or not the RPC eddy current data has been reanalyzed using terrain mapping of the data.
If terrain plots were not routinely used during the outage and your data has not been reanalyzed with terrain mapping of the data, discuss your basis for not reanalyzing your previous RPC data in light of the findings at Maine Yankee.
Discuss whether terrain plots will be used to analyze the RPC eddy current data at locations susceptible to circumferential cracking during your next steam generator tube inspection (i.e., the next inspection per your GL.95-03 response).
lp December 19, 1995
'v I,
r Dr. Robert C. Mecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649 r
SUBJECT:
R.
E.
GINNA NUCLEAR POWER PLANT REQUEST FOR ADDITIONAL "INFORMATION (RAI) CIRCUMFERENTIAL CRACKING OF STEAM GENERATOR TUBES (TAC NO. H92244)
Dear Dr. Mecredy:
The NRC staff has reviewed your Hay 5,
- 1995, response to Generic Letter 95-03, "Circumferential Cracking of Steam Generator Tubes," for your plant, and has identified several areas where additional information is needed.
Enclosed is an RAI for R.
E. Ginna Nuclear Power Plant.
Please provide your response within 30 days from the receipt of this RAI.
This request is within the original reporting burden for information collection of 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> covered by the Office of Management and Budget clearance number 3150-0011, which expires July 31, 1997.
Sincerely, Original signed by:
Docket No. 50-244
Enclosure:
Request for Additional Information cc w/encl:
See next page Distribution:
Docket File LMarsh PUBLIC SLittle PDI-1 Reading AJohnson SVarga
DOCUMENT NAME:
G: ~I,G INNA~I,G192244. RAI Allen R. Johnson, Project Manager Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation LDoerflein, RGN-I JStrosnider KKarwoski DHcDonald
- WLazarus, RGN-I To receive a copy of this document, indicate in the box: "C" ~ Copy without enclosures E"
Copy with enclosures "N" ~ No copy OFFICE NAME PM: PDI-1 LA:PDI-SLittl AJohnson:smm D:PDI-L arsh DATE 12/ I 95 12/
95 12/
/95 OFFICIAL RECORD COPY 12/
/95 12/
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