IR 05000302/2005009

IR 05000302-05-09, on 08/22-25/2005, Crystal River Unit 3 - NRC Annual Sample of Problem Identification and Resolution

ML052790006
Person / Time
Site:	Crystal River
Issue date:	09/29/2005
From:	Lesser M Division of Reactor Safety II
To:	Young D Florida Power Corp, Progress Energy Florida
References
IR-05-009
Download: ML052790006 (15)

Text

ber 29, 2005

SUBJECT:

CRYSTAL RIVER UNIT 3 - NRC ANNUAL SAMPLE OF PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000302/2005009

Dear Mr. Young:

On August 25, 2005, the US Nuclear Regulatory Commission (NRC) completed an inspection of a problem identification and resolution sample at your Crystal River Unit 3. The enclosed inspection report documents the inspection findings, which were discussed on August 24, 2005, and September 26, 2005, with members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents one NRC-identified violation of very low safety significance (Green) and one NRC-identified Severity Level IV violation. However, because of their very low safety significance and because each was entered into your corrective action program, the NRC is treating these violations as non-cited violations (NCV) consistent with Section VI.A of the NRC Enforcement Policy. If you contest any NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, NRC Region II; The Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at the Crystal River Unit 3 site.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Mark S. Lesser, Chief Engineering Branch 3 Division of Reactor Safety Docket No.: 50-302 License No.: DPR-72

Enclosure:

(See next page)

FPC 2 Enclosure: Inspection Report 05000302/2005009 w/Attachment: Supplemental Information

REGION II==

Docket No: 50-302 License No: DPR-72 Report No: 05000302/2005009 Licensee: Progress Energy Company Facility: Crystal River Nuclear Plant, Units 3 Location: 15760 West Power Line Street Crystal River Florida, 34428 Dates: August 22 through August 25, 2005 Inspectors: J. Blake, Consultant K. Karwoski, NRR/Division of Engineering Approved by: Mark Lesser, Chief Engineering Branch 3 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS IR 05000302/2005009; 08/22/2005 - 08/25/2005; Crystal River, Unit 3; Inspection of Problem Identification and Resolution - Annual Sample.

The inspection was conducted by a Region II consultant and an NRR Senior Technical Expert for Steam Generators.

NRC Identified Findings Cornerstone: Barrier Integrity C Green The inspectors identified a Non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for inadequate corrective action during review of the results of the refueling outage 12 steam generator Tube End Crack inspections. As a result, Crystal River 3 operated with the calculated leakage exceeding the Technical Specification leakage limit. The licensee entered this condition into their corrective action program.

The finding is more than minor because it was associated with steam generator tube integrity and affected the barrier integrity cornerstone, and if left uncorrected, a more significant safety concern could occur if appropriate corrective actions were not applied to unexpected results found during steam generator inspection activities. This finding represents a cross cutting aspect of problem identification and resolution. The finding was of very low safety significance because it did not result in loss of structural integrity of the steam generators, the small increase in estimated leak rate under main steam line break accident scenarios would not have any significant effect on core damage frequency or large early release frequency, and the contained location of flaws in the tubes makes it impossible to cause spontaneous tube ruptures.

C Severity Level IV The inspectors identified a Non-cited violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information, for several examples of inaccuracies and incomplete information in required reports and correspondence. The licensee entered this condition into their corrective action program.

This violation was assessed using traditional enforcement because it impacted the regulatory process. The issue is more than minor because the NRC relies on complete and accurate information to reach conclusions concerning the allowable time between steam generator inspections. It was determined to be a Severity Level IV violation because it was not willful, the technical issue associated with the incomplete and inaccurate information was of very low safety significance, and the NRC had not yet made a regulatory decision based on the information.

Report Details 4. OTHER ACTIVITIES (OA)

4OA2 Identification and Resolution of Problems Effectiveness of Problem Identification and Resolution a. Inspection Scope The inspectors reviewed the implementation of the licensees, NRC approved, alternate repair criteria (ARC) for indications found at the tube ends of the once-through steam generators (OTSG). This repair criteria is referred to as the tube-end crack (TEC) ARC, and it allows the licensee to leave flaws near the tube ends in-service provided that the calculated, accident-induced, primary-to-secondary leakage from those flaws meets the design/licensing basis requirements. Crystal River 3 began implementing the TEC ARC in their 1999 refueling outage (designated 11R). Subsequent refueling outages in which this ARC was implemented were in 2001 (12R) and 2003 (13R).

The inspectors reviewed the licensees implementation of the TEC ARC through discussions with responsible licensee personnel and review of the following documents:

• Condition monitoring assessments for the 11R, 12R and 13R OTSG inspections.

(Condition monitoring assessments verify that the as-found condition of the OTSGs are within acceptance limits)

• Operational assessments for the 11R, 12R and 13R OTSG inspections.

(Operational assessments are calculations performed to ensure that the acceptance limits will be met at the end of the next operating cycle.)

• Calculations performed by the licensee and contractors to support the conclusions in the condition monitoring and operational assessments.

• Licensee program documents and procedures for the inspection of the OTSGs

• Documentation of OTSG inspection results (including licensee correspondence with NRC concerning the results of the OTSG inspections.)

• Licensee nonconformance reports involving OTSG inspection activities This review included a comparison of the as-found OTSG conditions reported in the condition monitoring assessments of the 12R and 13R outages with the conditions predicted in the operational assessments of the 11R and 12R outages.

b. Findings Introduction: A Green NCV was identified when it was determined that corrective actions taken during the 12R steam generator inspections did not prevent the calculated accident-induced, primary-to-secondary leakage of the steam generators from being in excess of accepted leakage limits when the unit shut down for the 13R inspections.

Description: During a review of documentation for the 11R, 12R, and 13R refueling outage steam generator TEC ARC inspections, the inspectors determined that appropriate corrective actions were not taken during analysis of the 12R inspection data and calculations, (when an unexpected number of tube end cracks were found during

the OTSG inspections of 12R,) and as a result, when the unit shut down for the 13R refueling outage inspections, the steam generator as-found, calculated leakage exceeded the established leakage limits.

Two types of calculations are performed for assessing steam generator, accident-induced, primary-to-secondary leakage. One of the calculations is referred to as the condition monitoring assessment. This calculation verifies that the as-found condition of the steam generator is within acceptance limits. The second calculation is referred to as the operational assessment. This calculation is performed to ensure that the acceptance limits will be met at the end of the next operating cycle. In other words, the condition monitoring assesses the as-found condition and the operational assessment is forward looking to ensure that at the next inspection, the as-found condition will meet the acceptance limits.

The Crystal River TEC ARC was first used in the inspections conducted during the 11R outage. For the 11R steam generator inspections, the licensee made an assumption that additional TEC indications would be found during the 12R inspections due to the growth of existing indications and possible improvements in detection capabilities. This number was referred to as a probability of detection (POD) factor that was added to the as-left calculated leakage for inclusion in the operational assessment.

The 12R condition monitoring assessment reported significantly more TEC indications than predicted by the POD factor included in the 11R operational assessment. Despite this apparent discrepancy, the 12R operational assessment used the same POD factor (from the 11R assessment) for the prediction of calculated leakage expected at the 13R inspection. As a result, the licensee only repaired enough tubes to reduce the calculated leakage to a number slightly less than the allowable leakage less the POD factor. (A licensee review after the 13R inspection showed that if they had more effectively accounted for the increase in TEC indications, instead of just relying on the POD factor, they would have predicted that the as-found calculated leakage at 13R would exceed the established acceptance limits, and would have repaired more tubes prior to completion of the 12R inspections.)

Analysis: The finding was more than minor because it was associated with steam generator tube integrity and affected the barrier integrity cornerstone objective to provide reasonable assurance that the reactor coolant system protect the public from releases caused by accidents. If left uncorrected, the plant could operate with a larger value of calculated steam generator tube end crack leakage than allowed for by the license. The TEC ARC issue was determined to be of very low safety significance because: 1) the condition did not result in a loss of structural integrity of the steam generators, 2) the small increase in the estimated leak rate under main steam-line break accident conditions would not have any significant effect on the probability of core damage or large early release frequency during that design-basis accident or similar operational transients, and 3) the constrained location of these flaws in the tube ends makes it impossible for them to cause spontaneous tube ruptures. The finding also represents a cross-cutting aspect of problem identification and resolution.

Enforcement: 10 CFR 50, Appendix B, Criterion XVI, states in part that conditions adverse to quality shall be promptly identified and corrected. The Crystal River Unit 3

Steam Generator Integrity Program requires, in part, that during the review of inspection results, as a part of the integrity assessment process, the previous outage operational assessment is to be reviewed for accuracy, and if found to be inaccurate, corrective actions are to be taken to determine the cause and to update the input variables for the next operational assessment. Contrary to the above, on August 25, 2005, it was determined that corrective actions were not adequate in that this corrective action step was not done during the review of 12R steam generator inspection results, and as a result, the 13R as-found condition of the steam generators, relative to the tube end cracking alternate repair criteria, were outside the NRC approved, calculated leakage limits.

When the steam generators were found to be outside the approved calculated leakage limits during the 13R outage, the conditions were entered into the licensees corrective action process, and additional tubes were repaired to provide more margin between the alternate repair criteria, as-left calculated leakage and the approved calculated leakage limits. Because this failure to conduct adequate corrective actions is of very low safety significance and has been entered in the licensees corrective action program, this violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000302/2005009-01, Failure to Conduct Adequate Corrective Actions During Review of Steam Generator Inspection Results During 12R Refueling Outage Inspections.

4OA3 Event Followup (Closed) Licensee Event Report (LER)050000302/2004-04-00, NUREG-1022 Clarification Required Reporting of Previous Steam Generator Tube Inspection Results.

a. Inspection Scope In September 2004, NRC published a Notice of Clarification to Steam Generator Tube Integrity Guidelines in NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73. The NRC directed licensees to consider the results of the previous steam generator tube inspections, including the structural integrity criteria and leakage criteria, against the corrected reporting guidelines. Crystal River Unit 3 determined that the as-found steam generator projected leakage value for Steam Line Break (SLB), which exceeded the leak rate limit for Refueling Outage 13 (13R) was reportable per the corrected NUREG-1022 guidance. The inspectors reviewed this LER in conjunction with the corrective action inspection reported in 4OA2, above.

b. Findings The inspectors identified NCV 05000302/2005009-01, Failure to Conduct Adequate Corrective Actions During Review of Steam Generator Inspection Results During 12R Refueling Outage Inspections during the inspection reported in 4OA2, above. The corrective actions discussed in that section of the report were directly responsible for the conditions which resulted in the need for the LER, therefore the LER is considered to be closed by the NCV 05000302/2005009-01.

4OA5 Other Activities Completeness and Accuracy of Information a. Inspection Scope The inspectors reviewed correspondence between the licensee and NRC concerning the reporting of the results of the OTSG inspections. The focus of the inspection concerned correspondence and reports addressing the results of the implementation of the tube-end cracking (TEC), alternate repair criteria (ARC) which NRC approved for use starting with the 11R refueling outage inspections.

b. Findings Introduction: A Severity Level IV NCV was identified when it was determined that some correspondence and reports provided to the NRC, concerning the results of TEC ARC steam generator inspections during the 11R, 12R, and 13R outages, contained incomplete and/or inaccurate information.

Description: To ensure the condition of the steam generator tubes will remain acceptable over the course of the next operating cycle, the licensee provides an operational assessment which among other things, projects the amount of accident-induced leakage they would expect at the end of the next operating cycle. This assessment of the projected condition of the tubes at the end of the next operating interval includes (1) an assessment of the amount of leakage from the indications left in service (which may differ from the number of indications detected since some indications may have been repaired), and (2) an assessment of the amount of leakage from indications which may not have been detected during the outage or which may initiate during the cycle (a probability of detection adjustment referred to as leakage from undetected indications). During the review of licensee reports of inspection results, reviewers from the Office of Nuclear Reactor Regulation (NRR) found several examples of incomplete and/or inaccurate information which required extensive interaction and clarification with the licensee prior to acceptance by NRC. The problems identified by the NRR review staff were reviewed with the licensee during the inspection.

Pertinent examples of issues discussed were as follows:

As a result of reviewing documentation pertaining to the 11R outage (1999), NRC identified the following:

• On November 5, 1999, in Crystal River Unit 3 - Special Report 99-03: Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4 and Results of OTSG Tube Inspections that Fall into Category C-3." The licensee reported to the NRC that the condition monitoring assessment clarified that the structural and leakage integrity limits for the steam generators in the aggregate were not exceeded during the previous cycle. However, the licensees calculation showed that the amount of primary-to-secondary leakage under postulated accident conditions exceeded the 0.856 gpm limit when the NRC approved methodology was used. The licensees conclusion that the condition

monitoring limits for accident induced leakage were met was based on a methodology different than that approved by the NRC in the license amendment for the TEC ARC. The fact that an unapproved methodology was used was not reported and is considered incomplete information.

As a result of reviewing documentation pertaining to the 12R outage (2001), the staff identified the following:

• In a letter to the NRC dated October 19, 2001, Crystal River Unit 3 - Special Report 01-01, Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4", the licensee reported the condition monitoring accident induced leakage in steam generator A as 0.564 gpm. In a letter to the NRC dated January 22, 2002, Crystal River Unit 3 - Special Report 02-01: Results of the Once Through Steam Generator (OTSG) Tube Inservice Inspection Conducted During Refueling Outage 12," this value was incorrectly reported as 0.504 gpm. After the NRC pointed out the mis-match in values, the licensee clarified that 0.564 gpm was the correct value in a letter dated December 2, 2002, Crystal River Unit 3 - Response to Request for Additional Information Regarding Fall 2001 Steam Generator Inspection." This is considered inaccurate information.

• In a letter to the NRC dated January 22, 2002, Crystal River Unit 3 - Special Report 02-01: Results of the Once Through Steam Generator (OTSG) Tube Inservice Inspection Conducted During Refueling Outage 12," the licensee reported the leakage from indications left in service as the ARC (alternate repair criteria) leakage limit. The term leakage limit is used to reflect the maximum amount of leakage that can be tolerated consistent with the plants design and licensing basis. The terminology was corrected in a letter dated December 2, 2002, Crystal River Unit 3 - Response to Request for Additional Information Regarding Fall 2001 Steam Generator Inspection." This is considered inaccurate information.

• In the letter dated December 2, 2002, Crystal River Unit 3 - Response to Request for Additional Information Regarding Fall 2001 Steam Generator Inspection," the licensee reported the condition monitoring leakage for TEC indications in 12R as 0.56 gpm. As a result of reviewing the actual calculations, the NRC identified that the wrong values were reported to the NRC. The actual values were 0.57 gpm for OTSG A and 0.81 gpm for OTSG B. This is considered inaccurate information.

• The licensee used an incorrect table in determining the amount of leakage from TEC indications. This was identified by the licensee; however, the revised values were not reported to the NRC until after the NRC noticed (in 2005) that the values were not consistent with previous submittals. In addition, after using the correct values, the licensee determined that they had exceeded their design and licensing basis limit on the amount of accident induced leakage during 12R.

This fact was not highlighted to the NRC staff rather it was discovered based on a review of the information provided. (If the licensee had identified that they exceeded the accident induced leakage limit in 12R, they may have been able to

prevent exceeding the accident induced leakage limit during 13R.) This is considered incomplete information.

As a result of reviewing documentation pertaining to the 13R outage (2003), the staff identified the following:

• In Appendix 5 to the letter dated January 27, 2004, Crystal River Unit 3 - Special Report 04-01: Results of the Once-Through Steam Generator (OTSG) Tube Inservice Inspection Conducted During Refueling Outage 13" the licensee incorrectly included some tubes with TEC indications that are in-service but had been repaired (rerolled) during the 13R outage. A revised listing of tubes was provided in Attachment B to a letter dated March 30, 2005, Crystal River Unit 3 -

Response to NRC Request for Additional Information Regarding Once-Through Steam Generator Tube Inservice Inspection Conducted During Refueling Outage 13." This is considered inaccurate information.

• In the letter to the NRC dated March 30, 2005, Crystal River Unit 3 - Response to NRC Request for Additional Information Regarding Once-Through Steam Generator Tube Inservice Inspection Conducted During Refueling Outage 13,"

the licensee defined how new leakage was to be calculated; however, the values provided for new leakage in that letter did not appear to be calculated consistent with the definition. Subsequently, the NRC staff determined (through on-site interview during this inspection) that the values reported in the March 30, 2005 letter used a different definition of new leakage than that reported. This is considered incomplete information.

As a result of discussions with the licensee concerning the inconsistencies and errors discussed above, the licensee initiated a Nuclear Condition Report, NCR 20050823, OTSG Submittal Report Quality concerning the need for independent or third party review of OTSG information.

Analysis: The issue is more than minor because the NRC relies on complete and accurate information to reach conclusions concerning the allowable time between steam generator inspections. The NRCs NRR staff reviews licensee steam generator reports and assessments to provide confirmation that the facility can safely run until the next scheduled steam generator inspection, without having to shut down for mid-cycle inspections. The incomplete and/or inaccurate information provided to NRC concerning the Crystal River TEC ARC issue had an impact on the NRCs ability for oversight of licensed activities because extensive review and interaction was required to attain complete and accurate information.

The issue was determined to be a Severity Level IV Violation because it was not willful, the technical issue associated with the incomplete and inaccurate information was of very low safety significance, and the NRC had not yet made a regulatory decision based on the information.

Enforcement: 10 CFR 50.9, Completeness and Accuracy of Information, Paragraph (a), requires in part, that information provided to the Commission ... shall be complete and accurate in all material respects. Contrary to the above, as confirmed during the

on-site inspections on August 22-25, 2005, some reports and letters provided to the NRC associated with the results of steam generator examinations were incomplete or inaccurate with respect to information necessary for the NRC staff to complete the reviews of the information submitted.

Because this failure to provide complete and accurate information is a Severity Level IV violation and has been entered in the licensees corrective action program, this violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000302/2005009-02, Completeness and Accuracy of Information Provided to the NRC Concerning Steam Generator Inspection Results.

4OA6 Management Meetings The inspectors presented the inspection results to Mr. and other members of licensee management at the conclusion of the inspection on August 25, 2005 and in a telephone call to Mr. S. Powell on September 26, 2005. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.

SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee personnel M. Annacone, Engineering L. Cecilia, Licensing D. Herrin, Licensing T. Hobbs, NAS P. Peterson, NDE Level III S. Powell, Licensing D. Roderick, DSO S. Stewart, Steam Generator Engineer D. Taylor, Financial T. Williams, System Engineering D. Young, Site Vice President NRC personnel T. Morrissey, Senior Resident Inspector ITEMS OPENED AND CLOSED Opened and Closed 05000302/2005009-01 NCV Failure to Conduct Adequate Corrective Actions During Review of Steam Generator Inspection Results During 12R Refueling Outage Inspections. 05000302/2005009-02 NCV Completeness and Accuracy of Information Provided to the NRC Concerning Steam Generator Inspection Results.

Closed 050000302/2004-04-00 LER NUREG-1022 Clarification Required Reporting of Previous Steam Generator Tube Inspection Results.

Attachment

PARTIAL LIST OF DOCUMENTS REVIEWED Procedures Crystal River Unit 3 Plant Operating Manual, Steam Generator Integrity Porgram SP-305, OTSG Inservice Inspection Licensee Nonconformance Reports AR 00107734, OTSG B Lower Tube End Crack Identified, 10/15/2003 AR 00108677, Calculation Error for OTSG TEC Leakage in 12R Outage, 10/24/2003 AR 00109383, OTSG Condition Monitoring MSLB Leakage Results AR 00111677, OTSG Projected MSLB leakage exceeds OTSG Maintenance Rule Performance Criteria, 11/23/2003 AR 00138308, Reversal of NCR 109383 Reportability Determination AR 00157080, Inconsistencies in OTSG Report Information Correspondence Letter from NRC to Florida Power Corporation, Crystal River Unit 3 - Issuance of Amendment Regarding Alternate Repair Criteria for Steam Generator Tubing (TAC No. MA5395). dated October 1, 1999, Letter to the NRC, Crystal River Unit 3 - Special Report 99-03: Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4 and Results of OTSG Tube Inspections that Fall into Category C-3." dated November 5, 1999, Letter to the NRC, Crystal River Unit 3 - Special Report 01-01, Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4." dated October 19, 2001, Letter to the NRC, Crystal River Unit 3 - Special Report 02-01: Results of the Once Through Steam Generator (OTSG) Tube Inservice Inspection Conducted During Refueling Outage 12."

dated January 22, 2002, Letter to the NRC, Crystal River Unit 3 - Response to Request for Additional Information Regarding Fall 2001 Steam Generator Inspection." dated December 2, 2002, Letter to the NRC, Crystal River Unit 3 - Special Report 03-01: Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4." dated October 31, 2003, Letter to the NRC, Crystal River Unit 3 - Special Report 04-01: Results of the Once-Through Steam Generator (OTSG) Tube Inservice Inspection Conducted During Refueling Outage 13."

dated January 27, 2004, Letter from NRC to Florida Power Corporation,Crystal River Unit 3 - Summary of Conference Calls with Florida Power Corporation Regarding the Fall 2003 Steam Generator Inspection (TAC No. MC0965). dated February 17, 2004, Attachment

Letter from NRC to Florida Power Corporation, Review of Crystal River, Unit 3, Steam Genrator Tube Inservice Inspection Summary Reports From the Fall 2001 Outage (TAC No.

MC0965). dated April 13, 2004, Letter to the NRC, Crystal River Unit 3 - Response to NRC Request for Additional Information Regarding Special Report 03-01 (TAC No. MC1853)." dated August 10, 2004, Letter to the NRC, Crystal River Unit 3 - Response to NRC Request for Additional Information Regarding Special Report 03-01 (TAC No. MC1853)." dated September 9, 2004, Letter from the NRC to Florida Power Corporation, Request for Additional Information Regarding the Crystal River Unit 3 - Special Report 03-01: Once Through steam Generator Notifications Required Prior to Mode 4, and Special Report 04-01: Results of the Once-Through Steam Generator Tube Inservice Inspection Conducted During Refueling Outage 13 (TAC No MC1853). dated October 6, 2004 Letter to the NRC, Licensee Event Report 50-302/2004-004-00." dated November 22, 2004 Letter to the NRC, Crystal River Unit 3 - Response to NRC Request for Additional Information Regarding Once-Through Steam Generator, Special Reports 03-01 and 04-01." dated November 24, 2004, Letter to the NRC, Crystal River Unit 3 - Response to NRC Request for Additional Information Regarding Once-Through Steam Generator Tube Inservice Inspection Conducted During Refueling Outage 13." dated March 30, 2005 Letter to the NRC, Crystal River Unit 3 - Revised Response to NRC Request for Additional Information Regarding Once-Through Steam Generator Tube Inservice Inspection Conducted During Refueling Outage 13." dated May 20, 2005, Letter to the NRC, Crystal River Unit 3 - License Amendment Request #290, Revision 1, Probabilistic Methodology to Determine the Contribution to Main Steam Line Break Leakage Rates for the Once-Through Steam Generator from the Tube End Crack Alternate Repair Criteria." dated August 12, 2005 Attachment