ML050260021
| ML050260021 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/01/2005 |
| From: | Mozafari B NRC/NRR/DLPM/LPD2 |
| To: | Young D Progress Energy Florida |
| mozafari B, NRR/DLPM, 415-2020 | |
| References | |
| TAC MC1176, TAC MC1853 | |
| Download: ML050260021 (6) | |
Text
February 1, 2005 Mr. Dale E. Young, Vice President Crystal River Nuclear Plant (NA1B)
ATTN: Supervisor, Licensing and Regulatory Programs 15760 W. Power Line Street Crystal River, Florida 34428-6708
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING CRYSTAL RIVER UNIT 3 - ONCE-THROUGH STEAM GENERATOR TUBE INSERVICE INSPECTION CONDUCTED DURING REFUELING OUTAGE 13 (TAC NOS.
Dear Mr. Young:
By letters dated October 31, 2003 (ML033090110), January 27, 2004 (ML040350037),
August 10, 2004 (ML042320561), September 9, 2004 (ML04042710359), November 22, 2004 (ML043340228), and November 24, 2004 (ML043350045), Florida Power Corporation, the licensee for Crystal River Unit 3 Nuclear Power Plant (also doing business as Progress Energy, Florida), submitted information pertaining to the steam generator tube inspections at Crystal River Unit 3 from the Fall 2003 refueling outage. Additional information concerning these inspections was summarized by the NRC staff in a letter dated February 17, 2004 (ML040490002).
Several of the letters referenced above contain licensee responses to NRC staff questions.
Based on a review of these responses, we have determined that further information is needed to complete our review because the information provided by the licensee did not sufficiently address the original questions. The additional information needed is discussed in the enclosed request for additional information (RAI) that was e-mailed to your licensing staff on January 24, 2005. As discussed in the January 25, 2005, teleconference, your staff agreed to respond within 30 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, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosure:
As stated cc w/enclosure:
See next page
February 1, 2005 Mr. Dale E. Young, Vice President Crystal River Nuclear Plant (NA1B)
ATTN: Supervisor, Licensing and Regulatory Programs 15760 W. Power Line Street Crystal River, Florida 34428-6708
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING CRYSTAL RIVER UNIT 3 - ONCE-THROUGH STEAM GENERATOR TUBE INSERVICE INSPECTION CONDUCTED DURING REFUELING OUTAGE 13 (TAC NOS.
Dear Mr. Young:
By letters dated October 31, 2003 (ML033090110), January 27, 2004 (ML040350037),
August 10, 2004 (ML042320561), September 9, 2004 (ML04042710359), November 22, 2004 (ML043340228), and November 24, 2004 (ML043350045), Florida Power Corporation, the licensee for Crystal River Unit 3 Nuclear Power Plant (also doing business as Progress Energy, Florida), submitted information pertaining to the steam generator tube inspections at Crystal River Unit 3 from the Fall 2003 refueling outage. Additional information concerning these inspections was summarized by the NRC staff in a letter dated February 17, 2004 (ML040490002).
Several of the letters referenced above contain licensee responses to NRC staff questions.
Based on a review of these responses, we have determined that further information is needed to complete our review because the information provided by the licensee did not sufficiently address the original questions. The additional information needed is discussed in the enclosed request for additional information (RAI) that was e-mailed to your licensing staff on January 24, 2005. As discussed in the January 25, 2005, teleconference, your staff agreed to respond within 30 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, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosure:
As stated cc w/enclosure:
See next page Distribution:
PUBLIC EDunnington (Hard copy)
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OFFICIAL RECORD COPY REQUEST FOR ADDITIONAL INFORMATION REGARDING THE CRYSTAL RIVER UNIT 3 RESULTS OF THE ONCE-THROUGH STEAM GENERATOR TUBE INSERVICE INSPECTION CONDUCTED DURING REFUELING OUTAGE 13 (2003)
DOCKET NO. 50-302 By letters dated October 31, 2003 (ML033090110), January 27, 2004 (ML040350037), August 10, 2004 (ML042320561), September 9, 2004 (ML04042710359), November 22, 2004 (ML043340228), and November 24, 2004 (ML043350045), Florida Power Corporation, the licensee for Crystal River Unit 3 Nuclear Power Plant,(also doing business as Progress Energy Florida), submitted information pertaining to the steam generator tube inspections at Crystal River Unit 3 during the Fall 2003 refueling outage (designated 13R). Additional information concerning these inspections was summarized by the NRC staff in a letter dated February 17, 2004 (ML040490002).
Several of the above referenced letters contain licensee responses to requests for additional information made by the NRC staff. Based on a review of these responses, the NRC staff has determined that further information is needed to complete its review.
- 1. With respect to your accident induced leakage assessment, please provide the following:
a) a technical description of the methodology used to project the number and location (tubesheet radius) of tube-end-crack indications and the technical basis for this methodology (e.g., a benchmarking of this methodology based on previous inspection data). Include in this description, the actual values used.
b) the actual value of leakage from tube-end-crack indications based on the number of projected indications (from 1a above) and the existing NRC-approved leakage model.
c) a clarification of the number of tubes with tube-end-crack indications and the number of tube-end-crack indications. The staff notes that there appears to be a discrepancy in the reported number of tubes with tube-end-crack indications and the number of indications in both steam generators. For example, in steam generator A, 957 tubes were reported to have 1228 tube-end-crack indications in the October 31, 2003, letter; however, Appendix 5 to the January 27, 2004, letter indicates 1105 tubes in steam generator A contained tube end cracks. In addition, Table A-3 of the August 10, 2004, letter indicates 1119 tubes contained 1474 indications which does not appear (based on a cursory count) to match the number of tubes listed in Tables A-1 and A-2 (1099 tubes).
Enclosure 2.
In the November 24, 2004, response to RAI question 1d, it was indicated that indications are classified as being attributed to wear based on bobbin coil data. These wear indications are then compared to previous inspection data to determine if there is a change in signal characteristics. Rotating probe examinations are then performed at these locations of wear if the indications are new, the bobbin signal characteristics have changed, or there is no previous rotating probe data. Given that crack indications have been found at locations also affected by wear, please provide the technical basis for this approach. This technical basis should include the following:
a) a description of the bobbin coil eddy current data parameters used to distinguish wear from other degradation mechanisms (including intergranular attack and cracking); In particular, discuss whether other degradation mechanisms (e.g.,
intergranular attack and cracking) may also pass the test for being called wear based on screening the bobbin data.
b) the data supporting these parameters for screening wear from other forms of degradation.
c) a description of the criteria used to determine if the bobbin signal characteristics have changed and the basis for this change criteria. Please discuss whether operational data supports your screening criteria. For example, in the cases where cracking has been observed in wear scars, discuss whether the bobbin signals from these indications would have met your criteria for performing rotating probe examinations. If field wear scars with cracks would not have met your criteria for performing rotating probe examinations, discuss what corrective actions will be taken (including the basis for concluding that existing wear scars (with the potential for cracks being present) will have adequate integrity at the time of your next inspection).
3.
In the November 24, 2004, response to RAI question 2e, it was indicated that the difference between certain inspection numbers was due to the sleeves installed in the steam generators. In response to question 2b, it was indicated that 163 sleeves were installed in each steam generator at the start of the 2003 outage. Clarify the difference between the inspection numbers for steam generator B which is 159 rather than 163 (as it is for steam generator A).
4.
In the November 24, 2004, response to RAI question 3, the size of indications with the largest voltages were provided. Given that the largest voltage indication may not always correspond to the most severe indication (in terms of structural and leakage integrity),
confirm that, in this case, the largest voltage indications were the most severe indications detected. In the future it would be beneficial to delineate the location of each imperfection based on whether the indications are in the original roll transition, the original roll expanded region (other than the transition and the tube end), the re-roll upper or lower roll transition, the re-roll expanded region, the unexpanded portion of tube in the tubesheet region (not between re-rolls), the free span portion of the tube (outside the tubesheet) etc. In addition, it would be beneficial to specify the number of re-rolls (if any) were in the tube at the time of the inspection if the imperfections are in the tubesheet region.
5.
In the November 24, 2004, response to RAI question 5, you indicated that the obstructed sleeves were attributed to tube/sleeve end damage from previous loose parts and no indications were identified in the portion of the sleeve and tube examined.
Please discuss the following with respect to this finding:
a)
State whether the loose part was identified and removed from the steam generator (presumably from the primary side of the steam generator). Discuss the source of the part. If a part was not identified, discuss the basis for concluding the damage was from a loose part.
b)
Discuss the location of the obstruction (i.e., the portion of tube/sleeve extending above the upper tubesheet). If the obstruction was located in the portion of tube/sleeve within the tubesheet (including the clad), discuss how the part caused the obstruction.
c)
Given that the upper sleeve joint is a mechanical joint, discuss how you confirmed that the obstruction did not result in a weakening of the joint (i.e.,
pulling of the sleeve away from the parent tube) such that the sleeve could not meet its original design criteria. Provide the extent of the obstruction and the technical basis for your conclusion.
d)
Although no rejectable flaw-like signals were identified during the inspection of the portion of the sleeve that could be inspected, discuss whether an obstruction in another sleeved tube could have weakened the joint such that the sleeve could no longer meet its original design criteria.
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. Richard L. Warden 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