View source for 3F0698-33, Forwards Util Responses to Five Topics Discussed W/Nrc on 980624 Re LAR 228,Rev 0.LAR 228 Revised to Provide Addl Info as Result of Completion of Eddy Current Testing Tube End Anomalies Data re-analysis

{{Adams
| number = ML20236E977
| issue date = 06/30/1998
| title = Forwards Util Responses to Five Topics Discussed W/Nrc on 980624 Re LAR 228,Rev 0.LAR 228 Revised to Provide Addl Info as Result of Completion of Eddy Current Testing Tube End Anomalies Data re-analysis
| author name = Cowan J
| author affiliation = FLORIDA POWER CORP.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000302
| license number = 
| contact person = 
| document report number = 3F0698-33, 3F698-33, NUDOCS 9807010319
| package number = ML20236E979
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 6
| project = 
| stage = Other
}}

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Fisrida Power E "RW2
                                    ,,        ,   ET.1".*C. u. cen.72 June 30,1998 3F0698-33 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==Subject:==
Exigent License Amendment Request #228, Revision 1 Once Through Steam Generator Tube Surveillance Program

==References:==
: 1. FPC to NRC letter, 3F0698-28, dated June 18,1998, " Exigent License Amendment Request #228, Revision 0"
: 2. FPC to NRC letter,3F0698-25, dated June 16,1998, " Crystal River Unit 3 Review of Industry Operating Experience Regarding Tube End Anomalies"
: 3. FPC to NRC letter, 3F0598-08, dated May 18,1998, "An Operational Assessment of Steam Generator Tube Degradation at Crystal River Unit 3" l

==Dear Sir:==

In Reference 1, Florida Power Corporation (FPC) submitted Exigent License Amendment Request (LAR) #228, Revision 0. The purpose of this letter is to provide the NRC with LAR #228, Revision 1. LAR #228 has been revised to provide additional information as a
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result of the completion of eddy current testing (ECT) tube end anomalies data re-analysis.                           i This revision also provides FPC's responses to five topics discussed with the NRC staff during a telephone conference held on June 24,1998, regarding LAR #228, Revision 0.
[
Reference 2 informed the .NRC that a review of industry operating experience regarding steam generator tube end anomalies was being performed by FPC. Reference 2 also stated the review could result in a request for enforcement discretion or other regulatory action.                            After completing the review, FPC has concluded that the Crystal River Unit 3 (CR-3) Once Through Steam Generators (OTSGs) remain operable; however, additional work is required to define the pressure boundary for the tube-to-tubesheet roll joints.
Pending the outcome of this work to define the tube-to-tubesheet pressure boundary, FPC is                            !
requesting the NRC to continue its review of LAR #228 as revised herein. The previous conclusion of No Significant Hazards Consideration is not affected by this revision.
9007010319 900630 PDR     ADOCK 05000302e P                       PDR 1
l                                        CRYSTAL RIVER ENERGY COMPLEX: 15760 W. Power Line Street
* Crystal River, Florida 344284708 * (352)7954486 A Florida Progress Company I-L

U.S. Nuclear Regulatory Commission 3F0698-33 Page 2 of 3 '
                                  - Attachment A provides FPC's responses to five topics discussed with the NRC staff during the
                                  , telephone conference cited above. Attachment B provides the proposed ITS change in shaded
                                  . font, and Attachment C provides the proposed ITS Page 5.0-17 with the appropriate revision l-                         bar. Commitments made to the NRC in this submittal are provided in Attachment D.
3 If you have any questions regarding this suomittal, please contact Ms. Sherry Bernhoft, Manager, Nuclear Licensing at (352) 563-4566.
Sincerely, hb4                   ~
                                                                                                                                                                 \
John Paul Cowan Vice President                                                                                                               i Nuclear Operations l
l JPC/lve Attachments
: xc:     ' Regional Administrator, Region 11 Senior Resident Inspector NRR Project Manager i
1

s' U.S. Nuclear Regulatory Commission 3F0698-33
,                                                                 Page 3 of 3 i                              .
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STATE OF FLORIDA COUNTY OF CITRUS John Paul Cowan states that he is the Vice President, Nuclear Operations for Florida Power Corporation; that he is authorized _on the part of said company to sign and file with the Nuclear Regulatory Commission the information attr.: bed hereto; and that all such statements made and matters set forth theiein are 'true and' correc: to the -best of his knowledge, information, and belief.
A         - -- g John Paul Cowan                                                   I Vice Presid:nt                                                  :
Nuclear Operations Sw'orn to and subscribed before me this 3oOday of Ta Pl e_.                                      ,1998, by John Paul Cowan.                                                                                                       l Y   '
                                                                                                                                      /
Signature of Notary Public USA ANN MCBRIDE e    Notary Public. State of Florlds j
i l    My Comm. Exp. 0ct. 25,1999
                                                                                                                                  .            Comm. No. CC 505458 l
(Print, type, or stamp Commissioned Name of Notary Public) l l
f                                                                                                                      Personally                     Produced Known        I       -OR-       Identification                    >
l

l FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKFsT NO. 50-302/ LICENSE NO. DPR-72 ATTACHMENT A 1
EXIGENT LICENSE AMENDMENT REQUEST #228, REVISION 1 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM FPC'S RESPONSES TO FIVE TOPICS DISCUSSED %TTII TIIE NRC ON JUNE 24,1998 REGARDING LAR #228, REVISION 0 l
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U.S. Nuclear Regulatory Commission                                                Attachment A l            3F0698-33                                                                            Page 1 of 2 l
1 CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-77                                     l EXIGENT LICENSE AMENDMENT REQUEST #228, REVISION 1 ONCE TIIROUGli STEAM GENERATOR TUBE SURVEILLANCE PROGRAM FPC'S RESPONSES TO FIVE TOPICS DISCUSSED WITII TIIE NRC ON JUNE 24,1998 REGARDING LAR #228, REVISION 0 Topic #1 Preliminary calculations of accident induced leakagefrom tube endflaus estimated a leak rate rangingfrom 0.011 to 0.126 gallonsper minute (gpm). Provide a more definite bounding valuefor the leak rate and discuss how the value uns determined (e.g., mock-up testing, analytical          l calcidations or operating esperience). Also, discuss other modes of degradation that were          '
considered in determining the overall leak rate.                                                   '
The leak rate values were determined by a combination of mock-ups and analysis. The preliminary accident induced leak rate was based on a conservative calculation performed by Framatome Technologies, Inc. (FTI). The calculations yielded a cumulative end-of-cycle projected accident leakage of 0.011 gpm in 'he " A" Once Through Steam Generator (OTSG) and 0.125 gpm in the "B" OTSG due to the tube end anomalies (TEAS). Subsequent to the preliminary calculation, FTl has completed the review of the eddy current data and has refined the accident induced leakage value. The revised value for the " A" OTSG remains the same, and for the "B" OTSG is 0.029 gpm. This value was determined by using leak rates obtained from test samples during the tests performed by FTI for Arkansas Nuclear One (ANO). These tests measured leakage from axial Electro Discharge Machining (EDM) notches (0.25 inch x 0.005 inch) representing axial flaw.;,
and a roll joint containing a 360 circumferential severance representing circumferential indications. These tests were performed at 2,500 psi and room temperature, which was then adjusted for operating temperature. The leakage calculation indicates that the cumulative tube end anomaly accident leak rate, assuming all 381 indications in " A" OTSG and all 787 indications in
             " B" OTSG leak at end-of-cycle, is 0.011 gpm in " A" OTSG and 0.029 gpm in "B" OTSG.
This leak rate value from the TEAS will be added to FPC's previously submitted OTSG operational assessment leak rate determination provided in Reference 3. The resultant cumulative projected end-of-cycle leakage for the limiting case OTSG (the "B" OTSG is modeled as the most limiting generator), under main steam line break (MSLB) conditions, is 0.030 gpm.
Topic #2 What is the currentprimary-to-secondary leakage through each OTSG at Crystal River Unit 3?
The current total primary-to-secondary leakage through both OTSGs is less than one gallon per day.
w-_______

U.S. Nuclear Regulatory Commission                                                   Attachment A 3F%98-33                                                                                 Page 2 of 2
                    ' -'' To'pid#3
                                  ' Describe the nature of other indications identified in the re-analysis of eddy current testing (ECT)
{                                . datafor tube end anomalies.                                                                                      \
l No other indications were identified in the re-analysis of ECT data for tube end anomalies.                       l l
   'l Topic #4 Discuss uhether tubes that may be classified as defective, via tube end defects, nill be repaired during any maintenance outage ofsuficient duration prior to the nat refueling outage.
l
                                 - FPC is actively working with FTI to define the rolljoint pressure boundary. The results of a study originally requested by the B&W Owners Group will determine if any of the tubes with TEAS or                     i multiple end anomalies (MEAS) need to be reclassified as defective. . The analytical work is scheduled to be performed this summer with confirmatory laboratory testing to be conducted-through the remainder of 1998. FPC does not plan to take any action regarding plugging / sleeving
                                 - tubes with TEA indications until this work is completed.
l If any of the tubes are reclassified as defective, based on this study and testing, FPC will then                I schedule' repairs of the affected' tube ends either during Refueling Outage 11 (11R) or during an outage of sufficient duration if such an outage occurs prior to 11R. An outage of sufficient duration requires a' set of conditions that will result in placing the plant in a Cold Shutdown condition long enough to plan forithe tube plugging or repair activities. The activities would
                                 . include mobilization of the necessary (and available) manpower and equipment, drain down of the:                 j OTSGs, performing tube end inspections, plugging or repairing the affected tubes, and closeout ofL
                                 ' the OTSGs. FPC estimates that a 60-day outage would be required.
As' discussed during the June 24,1998,' telephone conference, another consideration for performing OTSG tube repairs prior to.llR is the number of ITS changes that would be involved. FPC is    .
currently planning to' submit ITS changes to revise the analysis of record for the small break Imss of Coolant Accident (LOCA), raise the low pressure engineered safeguards (ES) actuation setpoint, and to request permission to re-roll the'OTSG tubes. FPC's current schedule is to submit these ITS changes with sufficient time to allow for review and approval by the NRC prior to 11R, currently o                           ' scheduled for the fall of 1999. -
1 Topic #5                                                                                                         '
I Are you aware of the location of the leak idennfied at ANO during their 01SG bubble test?
                                 - Yes,'it was characterized as a circumferential crack located just below the seal. weld to cladding interface.
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