3F0598-08, Forwards Rev 0 to AES98033350-1-1, Operational Assessment SG Tube Degradation at Crystal River Unit 3. Results of Assessment Provide Reasonable Assurance That OTSG Tube Integrity Will Be Maintained During Current Operating Cycle

Forwards Rev 0 to AES98033350-1-1, Operational Assessment SG Tube Degradation at Crystal River Unit 3. Results of Assessment Provide Reasonable Assurance That OTSG Tube Integrity Will Be Maintained During Current Operating Cycle

ML20247Q147
Person / Time
Site:	Crystal River
Issue date:	05/18/1998
From:	Baumstark J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20247Q155	List: 3F0598-08, Forwards Rev 0 to AES98033350-1-1, Operational Assessment SG Tube Degradation at Crystal River Unit 3. Results of Assessment Provide Reasonable Assurance That OTSG Tube Integrity Will Be Maintained During Current Operating Cycle
References
3F0598-08, 3F598-8, NUDOCS 9805280216
Download: ML20247Q147 (7)

Text

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Fisrida Power f 5 "NA" 8?.Co*uM ue.onn May 18,1998 3F0598-08 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

An Operational Assessment of Steam Generator Tube Degradation at Crystal River Unit 3

References:

1. FPC to NRC letter, 3F1297-22, dated December 5,1997, "Special Report 97-05, Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4, and Complete Results of OTSG Tube Inservice Inspection Performed During the Current Outage (90-Day Report)"

2. FPC to NRC letter, 3F0498-25, dated April 22,1998, " Commitment Due Date Extension"

Dear Sir:

The purpose of this letter is to fulfill the commitment made by Florida Power Corporation I (FPC) to complete and submit an operational assessment for first span pit-like Intergranular Attack (IGA), in the "B" Once Through Steam Generator, within 90 days following plant restart (Reference 1). The due date for submittal of the operational assessment was l

, extended to May 28,1998 (Reference 2).

Attachment A is an introductory summary to "An Operational Assessment of Steam Generator Tube Degradation at Crystal River Unit 3" (Attachment B). The scope of the

, attached operational assessment exceeds the scope of the commitment made in Reference 1  ;

in that it also provides an assessment of freespan axial outside diameter stress corrosion l cracking (ODSCC/ IGA) and axial primary water stress corrosion cracking (PWSCC) at j upper roll expansion transitions. The results of the operational assessment provide reasonable assurance that the OTSG tube integrity will be maintained during the current operating cycle. } }

9805280216 980518 \

PDR ADOCK 05000302 .

i l P PDR l

CRYSTAL RNER ENERGY COMPLEX: 15760 W. Power Line Street

• Crystal River, Florida 344284708 * (362)7954486 '

A Florida Progress Company j

~

U.S. Nuclear Regulatory Commission

., 3F0598-08 Page 2 of 2 i

There are no new commitments established by this letter or its attachments. If you have any questions regarding this submittal, please contact Ms. Sherry Bernhoft, Manager, Nuclear Licensing, at (352) 563-4566.

Sincerely, f

d\ , J.

._ ... -&_ _D_ -

S. Bt umstark, Director Nuclear Engineering and Projects JSB/JBM/LVC xc: Regional Administrator, Region II Senior Resident Inspector NRR Project Manager Attachments:

A. Introductory Summary to Operational Assessment of Steam Generator Tube Degradation at Crystal River Unit 3 B. An Operational Assessment of Steam Generator Tube Degradation at Crystal River Unit 3

+ .

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 l

ATTACHMENT A INTRODUCTORY

SUMMARY

TO OPERATIONAL ASSESSMENT OF STEAM GENERATOR TUBE DEGRADATION i

AT CRYSTAL RIVER UNIT 3 l l

• , a

'U.S. Nuclear Regulatog Commission Attachment A l

. 3F0598-08 Page1of3 I Background I l

Florida Power Corporation (FPC) submitted License Amendment Request (LAR) #221, Revision 0, to the NRC on October 1,1997 (Reference 1). In this LAR, FPC proposed a specific methodology for monitoring long term growth of the "B" Once Through Steam Generator (OTSG) first span pit- (

like Intergranular Attack (IGA) indications. The submittal proposed implementing enhanced j reporting requirements requiring FPC to perform " operational assessments" to determine the i adequacy of the structural integrity of the OTSGs during future OTSG inspections. As a conservative measure, FPC has performed an OTSG operational assessment using the eddy current data obtained during the 1997 extended outage. The operational assessment is forward-looking and  ;

is intended to demonstrate, with reasonable assurance, that the tube integrity performance criteria will be maintained throughout the operating cycle until the next scheduled tube inspection.

Following the submittal of LAR #221, FPC submitted a combined Mode 4 and 90-day Special Report (Reference 2), outlining the tube inspections performed and the results obtained during the 1997 inspections. Included with this submittal was a detailed assessment of the indicated growth trends for the "B" OTSG first span pit-like IGA indications. FPC committed, in the 90-day Special Report, to submit an operational assessment for first span pit-like IGA within 90 days following plant restart.

Draft Regulatory Guide DG-1074, " Steam Generator Tube Integrity," September 1997, and NRC Generic Letter (GL) 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," were used as references during the development of the CR-3 operational assessment. In Reference 1, CR-3 proposed the following information be provided as part of operational assessments:

1. For first span "B" OTSG pit-like IGA, a summay of analyses performed to determine if l estimated leakage based on the projected end-of-cycle percent throughwall distribution exceeds the leak limit (determined from the licensing basis dose calculation for the main steamline break) for the next operating cycle.

2. For first span "B" OTSG pit-like IGA, a summary of analyses performed to determine if the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, i using the actual measured end-of-cycle) percent throughwall distribution exceeds 1x10-2,and j an assessment of the safety significance of the occurrence if exceeded.

The operational assessment (Attachment B) considers these degradation mechanisms:

I e Freespan Axial Cracking - One indication found in "A" OTSG during the 1997

)

inspection 1

• Upper Tubesheet Roll Transition Cracking - 51 indications found in each OTSG during the 1997 inspection

. "B" OTSG First Span Pit-Like IGA - Predominant form of degradation l

I 1

1

- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i

4

'U.S. Nuclear Regulato y Commission Attachment A

, 3F0598-08 Page 2 of 3 An operational assessment that addresses each of these mechanisms provides a conservative estimate of the condition of the OTSGs.

Results Attachment B provides a summary of the operational assessment performed for CR-3 based upon the 1997 eddy current inspection of the OTSGs. This assessrnent is based primarily on Draft Regulatory Guide DG-1074, " Steam Generator Tube Integrity," September 1997. The document demonstrates that the CR-3 OTSG tubing is expected to maintain the appropriate margins ofleakage and stmetural integrity for both normal and accident conditions until the end of the current fuel cycle.

\

FPC has collected a considerable amount of site specific data for the "B" OTSG first span pit-like IGA indications. These eddy current indications have been observed and monitored since 1990 (Reference 1). Reference 2 provides a detailed assessment of first span pit-like IGA indicated growth trends. Because FPC utilizes a 40% throughwall plugging criteria for the "B" OTSG first span pit-like IGA, a deterministic approach is utilized to project tube condition at the end of the cycle. This assessment demonstrates that the projected largest indication at the end of this cycle is well below 100% throughwall, thus leakage is expected to be zero (0) gallons per minute (gpm) for normal and accident conditions. In addition to this analytical determination, the in-situ pressure testing performed in 1996 (Reierence 3) and the burst testing performed on tubes pulled from the "B" OTSG in 1992 and 1994 (References 4 and 5) continue to bound the indications left insenice.

Thus, these indications are not expected to cause these tubes to burst or leak insenice during normal or accident conditions.

Due to the limited amount of site specific data for freespan axial and upper tubesheet roll transition indications, a probabilistic assessment for leakage and stmetural integrity was performed for these degradation mechanisms. Because '997 was the first year that CR-3 identified these degradation mechanisms, this assessment also relies upon data collected from other plants. The CR-3 eddy current inspection protocol does not allow the assignment of a throughwall percentage to these indications, and requires them to be plugged on detection. This approach is consistent with industry practice, especially during the discovery phase of a new degradation mechanism. The assessment of the freespan axial indication is based primarily on data collected from other OTSG plants, while the assessment of axial indications in the roll transition area (inner diameter oriented) utilizes a much broader database ofindustry experience. Future CR-3 inspections will provide an opportunity to compare site specific initiation, growth, and detection data of axial indications to industry values, and to adjust the operational assessment appropriately.

Attachment B demonstrates that for the axial indication observed in the upper bundle freespan region and the upper tubesheet roll transition area, the conditional probability of burst at accident conditions is below the Draft Regulatory Guide DG-1074 acceptable limit of 0.025, and the projected leakage rate is wel1 below the allowable limit of I gpm. Accident condition primary to secondary leakage is limited to 1 gpm, per the CR-3 Improved Technical Specifications (ITS) Bases and accident analyses. This value is also consistent with DG-1074, Section C.2.3.

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U.S. Nuctrar Regulatory Commission Attachm:nt A y

3F0598-08 Page 3 of 3 l

i To summarize the calculated results:

l. Conditional Probability of Probability of Leakage at 95/95 Projected Burst Accident Condition Accident Conditions Leak Rate (gpm)at Accident Conditions Freespan Axial- 0.0012 0.0026 0.000 Indication i UpperTubesheet Axial 0.0074 l 0.153 0.001 Indications l Cumulative 0.0086 0.156 0.001 DG-1974 Cumulative Less than 0.025 T- .

. Less than 1 gpm P~ *- Criteria .

i Operational leakage at CR-3 is limited to 150 gallons per day (gpd), in accordance with the CR ITS. This operational leakage limit was implemented on a permanent basis by License Amendment .

No.158, and it is consistent with the recommended value in GL 95-05. l Conclusion l

Based upon the attact.ed nonrational assessment, reasonable assurance is provided that the CR-3 OTSG tube integrily peAmance criteria will be maintained until the end of the current operating cycle. Subsequent eddy current inspections of the OTSGs will provide the opportunity to increase site specific data.. This will allow CR-3 to characterize site specific initiation, growth, and detection .

curves for axial indications. This new data will be used to enhance and update the site specific operational assessment model as CR-3 continues to ensure OTSG tube integrity.

References

1. FPC to NRC letter, 3F1097-21, dated October 1,1997, " License Amendment Request No. 221, Revision 0, "B" Once-Through Steam Generator Tube Surveillance Program."

2. FPC to NRC letter,3F1297-22, dated December 5,1997, "Special Report 97-05, Once-Through Steam Generator (OTSG) Notifications Required Prior to MODE 4, and Complete Results of OTSG Tube Inservice Inspection Performed During the Current Outage (90-Day Report)."

3. FPC to NRC letter,3F0496-04, dated April 8,1996, " Technical Specification Change Request No. 203, Revision 2, Notifications Required Prior to MODE 4, and Responses to Request for Additional Information."

4. FPC to NRC letter, 3F0494-09, dated April 19,1994, " Refuel 9 Inspection Plan for Once Through Steam Generators."

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5. - FPC to NRC letter, 3F0595-07, dated May 31,1995, "Once Through Steam Generator '

Inspection, Technical Specification Twelve-Month Report."

i I

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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT B 1 l

l AN OPERATIONAL ASSESSMENT OF STEAM GENERATOR TUBE DEGRADATION AT CRYSTAL RIVER UNIT 3 l

l l