ML061630437
| ML061630437 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/22/2006 |
| From: | Mozafari B Plant Licensing Branch III-2 |
| To: | Young D Progress Energy Florida |
| Mozafari B, NRR/ADRO/DORL, 415-2020 | |
| References | |
| TAC MC9562, TAC MD0220, TAC MD0357 | |
| Download: ML061630437 (5) | |
Text
June 22, 2006 Mr. Dale E. Young, Vice President Crystal River Nuclear Plant (NA1B)
ATTN: Supervisor, Licensing and Regulatory Programs 15760 W. Power Line Street Crystal River, FL 34428-6708
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE 2005 STEAM GENERATOR TUBE INSPECTIONS AT CRYSTAL RIVER UNIT 3 (TAC NOS.
MC9562, MD0220, MD0357)
Dear Mr. Young:
By letters dated December 3, 2005, February 15, and March 8, 2006, Florida Power Corporation (the licensee also doing business as Progress Energy - Florida) submitted information summarizing the results of the 2005 steam generator tube inspections at Crystal River Unit 3. These inspections were performed during the fourteenth refueling outage.
In addition to these reports, the U.S. Nuclear Regulatory Commission staff participated in a conference call concerning the 2005 steam generator tube inspections at Crystal River Unit 3.
This call is in the process of being summarized.
In order for the NRC staff to complete its review of the licensees reports, we request that the licensee provide responses to the enclosed questions. The additional information needed is discussed in the enclosed request for additional information (RAI) that was discussed with your licensing staff on June 8, 2006, during a teleconference. As discussed in the teleconference, your staff agreed to respond within 45 days of the date of this RAI.
If you have any questions regarding this matter, please contact me at (301) 415-2020.
Sincerely,
/RA/
Brenda L. Mozafari, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosure:
Request for Additional Information cc w/enclosure: See next page
ML061630437 OFFICE NRR/LPL2-2/PM NRR/LPL2-2/LA EMCB/SC*
NRR/LPL2-2/BC NAME BMozafari BClayton TBloomer (Acting)
By memo dated MMarshall (DPickett for)
DATE 6/21/06 6/20/06 5/3/06 6/22/06
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE CRYSTAL RIVER UNIT 3 RESULTS OF THE ONCE-THROUGH STEAM GENERATOR TUBE INSERVICE INSPECTION CONDUCTED DURING REFUELING OUTAGE 14 (2005)
DOCKET NO. 50-302 By letters dated December 3, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML053410277), February 15, 2006 (ADAMS Accession No. ML060550264), and March 8, 2006 (ADAMS Accession No. ML060740185), Florida Power Corporation (the licensee also doing business as Progress Energy - Florida) submitted information summarizing the results of the 2005 steam generator tube inspections at Crystal River Unit 3. These inspections were performed during the fourteenth refueling outage (14R).
In order for the U.S. Nuclear Regulatory Commissions staff to complete its review of the these reports, we request responses to the questions presented below.
1.
In your December 3, 2005 letter, you indicated that postulated steam line break accident-induced leakage based on your 2005 inspection results was less than that projected from your previous (2003) inspection results. Clarify whether this was true for each of the degradation mechanisms for which leakage was projected. If there were any instances in which the projected leakage (based on 2003 results) was less than the as-found leakage (based on 2005 results) for a specific degradation mechanism, please discuss what corrective actions were taken.
2.
On page 3 of 6 in your December 3, 2005, letter, you provided the in situ test pressures for several indications. Discuss the basis for these test pressures. For example, was the 5550 pounds per square inch test pressure based on a large break loss-of-coolant accident? In addition, confirm that these test pressures were based on your most limiting condition (e.g., a design basis accident with the appropriate safety factor or three times your normal operating pressure differential).
3.
On page 3 of 16 in your March 8, 2006, letter, you indicated that approximately 68 tubes were degraded. Confirm that all of these degraded tubes were a result of wear.
If not, please discuss the basis for sizing the indications.
4.
On page 10 of 16 in your March 8, 2006, letter, you indicated that as an additional check of the tube-end-cracking (TEC) leakage methodology, you projected the 14R leakage based on the 13R inspection results. Provide the data supporting your conclusion that the methodology over-predicted the as-found TEC leakage (e.g., new leakage, as-left leakage, etc.). Provide the data for both the upper and lower tubesheet in both steam generators.
Enclosure 5.
Clarify the statement on page 10 of 16 in your March 8, 2006, letter, where you indicate that the linear projection method is still considered adequate since the projected leakage for 15R is larger than the 14R as-found leakage. It would appear that a more appropriate comparison for determining the adequacy of the linear projection method would be comparing the leakage projection from the previous outage with the as-found leakage at the next outage.
6.
On page 13 of 16 in your March 8, 2006, letter, you indicated that one Alloy 600 welded plug, which was considered to be leaking, was removed and replaced. Discuss whether this leaking plug had adequate structural integrity including a summary of the basis for your conclusion.
7.
Regarding the in situ pressure test of tube A3-30, discuss the magnitude of the change in the size of the indication as a result of the pressure test.
8.
Regarding your condition monitoring assessment, discuss the sources of accident-induced leakage and the amount of leakage assigned to each of those sources.
9.
Clarify the number and size of indications found in the unexpanded region of the tube within the upper and lower tubesheet. Indicate which indications were detected with the bobbin probe, rotating probe, or both probes.
10.
Clarify the number and size of indications attributed to groove intergranular attack (IGA)/stress corrosion cracking.
11.
A number of volumetric indications (other than first span IGA and wear) were identified.
Discuss the cause of these indications. Confirm that all volumetric indications (other than first span IGA and wear) were plugged.
Mr. Dale E. Young Crystal River Nuclear Plant, Unit 3 Florida Power Corporation cc:
Mr. R. Alexander Glenn Associate General Counsel (MAC-BT15A)
Florida Power Corporation P.O. Box 14042 St. Petersburg, Florida 33733-4042 Mr. Jon A. Franke Plant General Manager Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, Florida 34428-6708 Mr. Jim Mallay Framatome ANP 1911 North Ft. Myer Drive, Suite 705 Rosslyn, Virginia 22209 Mr. William A. Passetti, Chief Department of Health Bureau of Radiation Control 2020 Capital Circle, SE, Bin #C21 Tallahassee, Florida 32399-1741 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 Mr. Craig Fugate, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 Chairman Board of County Commissioners Citrus County 110 North Apopka Avenue Inverness, Florida 34450-4245 Mr. Michael J. Annacone Engineering Manager Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, Florida 34428-6708 Mr. Daniel L. Roderick Director Site Operations Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, Florida 34428-6708 Senior Resident Inspector Crystal River Unit 3 U.S. Nuclear Regulatory Commission 6745 N. Tallahassee Road Crystal River, Florida 34428 Mr. Terry D. Hobbs Manager, Nuclear Assessment Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, Florida 34428-6708 David T. Conley Associate General Counsel II - Legal Dept.
Progress Energy Service Company, LLC Post Office Box 1551 Raleigh, North Carolina 27602-1551