View source for 3F0698-28, Application for Amend to License DPR-72,exigent LAR 228,rev 0,proposing one-time Exigent License Amend to Allow Operation W/Number of Indications Previously Identified as Tube End Anomalies & Multiple Tube End Anomalies in OTSG

{{Adams
| number = ML20249B264
| issue date = 06/18/1998
| title = Application for Amend to License DPR-72,exigent LAR 228,rev 0,proposing one-time Exigent License Amend to Allow Operation W/Number of Indications Previously Identified as Tube End Anomalies & Multiple Tube End Anomalies in OTSG
| 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 = 3F0698-28, 3F698-28, NUDOCS 9806220245
| package number = ML20249B265
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 12
| project = 
| stage = Request
}}

=Text=
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l         June 18,1998 3F0698-28 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==Subject:==
Exigent License Amendment Request #228, Revision 0 Once Through Steam Generator Tube Surveillance Program

==Reference:==
FPC to NRC letter, 3F0698-25, Crystal River Unit 3 Review of Industry Operating Experience Regarding Tube End Anomalies

==Dear Sir:==

Pursuant to 10 CFR 50.90, Florida Power Corporation (FPC) hereby submits Exigent License Amendment Request (LAR) #228, Revision 0, for an amendment to the Crystal River Unit 3 (CR-3) Operating License No. DPR-72. As part of this request, FPC is providing the License Amendment Request, No Significant Hazards Consideration Evaluation, and Environmental Impact Evaluation Pages (Attachment A), Proposed Technical Specification Change Pages, Shaded (Attachment B), and Proposed Technical Specification Change Pages, Revision Bars (Attachment C).
FPC proposes a one-time Exigent License Amendment to allow operation with a number of indications previously identified as tube end anomalies (TEA) and multiple tube end anomalies (MEA) in the CR-3 Once Through Steam Generator (OTSG) tubes. The duration of the proposed License Amendment is until CR-3's next refueling outage (llR), currently scheduled for fall 1999.
This proposed change may be necessary due to the potential condition of noncompliance with CR-3 Improved Technical Specification 5.6.2.10.4.b. Such a condition may result from confirmation of an ongoing re-analysis of eddy current testing (ECT) data, of indications previously identified as TEAS and MEAS in the upper roll expansion as now being within the pressure boundary of the tubes. In the above referenced letter, FPC informed the NRC staff of this in-progress review of CR-3 ECT data, and the potential for a request for enforcement discretion once the review of ECT data is completed the week of June 22,1998. The confirmed information will be conveyed to the NRC upon completion.
The Improved Technical Specifications (ITS) pages provided in Attachments B and C to this letter have been affected by previously proposed LAR #221, Revisions 0 and 1. Changes provided in Attachments B and C were made to the current ITS pages, as approved by License Amendment No.158. Once this License Amendment is issued, FPC will provide updated ITS                           1
_ pages for LAR #221.                          ,                                                               \   g
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* Crystal River, Flor 6da 344284708 (352)7954486 9906220245 990618                            vorwe Proeren company PDR      ADOCK 05000302 P                                PDR   .

U.S. Nuclear Regulatory Commission
                  ,            3F0698-28 Page 2 of 3 There are no new regulatory commitments contained in this letter or its attachments. If you -
have any questions regarding this letter, please contact Ms. Sherry Bernhoft, Manager, Nuclear
                              . Licensing at (352) 563-4566.
Sincerely, h Pd            =
                            ' John Paul Cowan Vice President Nuclear Operations JPC/lve Attachments xc:       Regional Administrator, Region 11 Senior Resident inspector NRR Project Manager I
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iw                      U.S. Nuclear Regulatory Commission -
3F0698-28
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Page 3 of 3 L
STATE OF FLORIDA' I                       COUNTY OF CITRUS '
L l
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 attached hereto; and that'all such statements made and matters set forth therein are true and correct to the' best of his knowledge, information, and belief.
                                                                                         $&e3 John Paul Cowan Vice President Nuclear Operations Sworn to and subscribed before me this I6b                   day of                             3h                            ,1998, by John Paal Cowan.
                                                                                               /
W Signature of Notary Public State of Florida LtSA ANN MCBRtDE
                                                                                                                                  ~ Notary Public. State of HorMe t                        ,j             My comm. Exp. 0ct. 25.1999 4WgF                                     Comm. No. CC 505455 (Print, type,Tr stamp Commissioned Name of Notary Public)
Produced Personally Known          I                       -OR-              Identification

I FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 ATTACHMENT A EXIGENT LICENSE AMENDMENT REQUEST #228, REVISION 0 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM License Amendment Request, No Significant Hazards Pages, Environmental Impact Evaluation Pages I
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I
                                                                  )
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U.S. Nuclear Regulatory Commi;sion                                                  Attachment A
       . 3F0698-28                                                                                 Page 1 of 8
                                                                                                                                 .)
i CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 EX1 GENT LICENSE AMENDMENT REQUEST #228, REVISION 0 ONCE THROUGil STEAM GENERATOR TUBE SURVEILLANCE PROGRAM LICENSE DOCUMENT INVOLVED: Improved Technical Specifications PORTIONS: Technical Specification 5.6.2.10.4.b, " Steam Generator (OTSG) Tube Surveillance Program''

==SUMMARY==
OF CHANGES:
Florida Power Corporation (FPC) is currently in the process of re-analyzing the most recent Once Through Steam Generators (OTSG) inservice inspection eddy current testing (ECT) data. This review was prompted as a result of operating experience from Arkansas Nuclear One (ANO) and Oconee Nuclear Station (ONS). This operating experience indicated the potential for some                                 J indications previously identified as Tube End Anomalies (TEAS) and Multiple Tube End Anomalies (MEAS) in the upper roll expansions to be identified as being within the pressure boundary of the tubes.
Final confirmation of the tube ECT data and location of the indications with respect to the pressure boundary is not expected to be complete until the week of June 22, 1998. However, FPC recognizes the potential may exist that, upon completion of this review, the steam generators may be determined to not be in compliance with Improved Technical Specification 5.6.2.10.4.b, thereby affecting the operability of the OTSGs.
Based on this potential condition of notenmpliance, assuming worst case re-analysis results, FPC is proposing a one-time Exigent License Amedment Request to allow operation with potential TEA and MEA indications in the pressure boundary of the tubes until Refueling Outage 11, currently                           l scheduled for the fall of 1999. FPC has performed an operational and leakage assessment which                            l l        concludes that operating with a number of TEAS and MEAS within the tubes' pressure boundary                              1 would not affect the apability of the OTSGs to perform their intended safety function during normal operation and postulated accident conditions.
CHANGE TO SPECIFICATION 5.6.2.10.4.b, " Steam Generator (OTSG) Tube Surveillance Program" i
Description of Specification Change Add new paragraph to improved Technical Specifications (ITS) 5.6.2.10.4.b, page 5.0-17, to read                          )
as follows:                                                                                                              j l
                                                                                                ._       _-_ _ _ _ __-_ _    A

m
\ .
U.S. Nuclear Regulatory Commission                                                 Attachment A
      . 3F0698-28                                                                             - Page 2 of 8 l
l                     ,       .
: b.        De OTSG shall be determined OPERABLE after completing the corresponding actions (plug or sleeve all tubes exceeding the plugging /steeving limit and all tubes containing through-wall cracks) required by Table 5.6.2.2 (and Table 5.6.2-3 jf the prevbions of Specification 5.6.2.10.2.d am utilized). Defective tubes may be mpaired in accordance with the B&W prvcess (or method) equivalent to the method decribed
                ' in Report BAW-2120P.
Here are a number of OTSG tubes that exceed the tube plugging / sleeving limit as a result of tube end anomalies. Dese tubes will le repaired in the next refueling outage (11R). An analysis has been performed which confirms that operability of CR-3 OTSGs will not be impacted with these tubes inservice until the next refueling outage (IIR).

===Reason for Request===
ITS 5.6.2.10.4.b requires that OTSG tubes exceeding the tube plugging / sleeving must be plugged or sleeved for the OTSGs to be considered operable. This proposed change is necessary due to the potential condition of noncompliance that would result from confirmation, by the ongoing re-analysis of ECT data, of indications previously identified as TEAS and MEAS in the upper roll expansion which could potentially be located in the pressure boundary of the tubes.
Justification for Request

===Background===
Based on information provided by ANO and ONS, Crystal River Unit 3 (CR-3) initiated Precursor Card 3C-98-2857 on June 9,1998, to document industry operating experience pertaining to in-service steam generator eddy current indications which may exceed the ITS plugging . limit. The indications in question are referred to as TEAS and MEAS. These designations have been used for identifying indications in the Babcock & Wilcox (B&W) OTSG tube ends which protrude past the tubesheet. These designations have historically been used only on indications which have been considered to be located outside the pressure boundary, and thus the indications could be left in service. If the indications were located in the perceived pressure boundary, the indications would be identified as Single Axial Indications (SAls), Single Circumferential Indications (SCis) or other conventional designation. As such, these indications would have been plugged in the 1997 outage.
In the summer of 1997, CR-3 performed an extensive ECT examination of both OTSGs. The scope of examination included inspecting 100% of the upper tubesheet (hot leg) roll transitions and roll expansions using a rotating coil inspection probe. The rotating coil probe included both pancake and Plus-Point coils to achieve maximum inspection sensitivity.
Numerous tubes were identified with TEAS and MEAS. At the time, the eddy current analysts used the B&W Owners Group (BWOG) industry protocol for identifying the tube pressure boundary as

                           ' U.S.- Nuclear Regulatory Commission                                                  : Attachment A' 3F0698-28                                                                                Page 3 of 8 I
up to and including the portion of data which indicates that the probe has left the carbon' steel
                           - portion of the tubesheet and is _in. the inconel clad region. Using this methodology and understanding of vbt defined the pressure boundary, 273 tubes in the "A" OTSG and 554 tubes in the "B" OTSG w. e wturned to service with TEAS or MEAS in this region of the tube.
However, since 1996, Framatome Technologies, Inc. (FTI) has been reviewing and defining what constitutes the pressure boundary for the tubes in the OTSG design. This effort has been driven by BWOG member utilities' desire to implement repair roll processes at the OTSG plants to keep -
tubes in service longer. In the spring of 1998, FTl identified, verbally, to the utilities that the pressure boundary (for a non-repaired tube) is defined as the portion of tube which extends from the primary side face of the inconel cladding on the upper tubesheet to the primary side face of the inconel cladding on the lower tubesheet. As a result, ANO-1 and ONS examined their OTSG 'r oll expansion regions more closely in spring 1998 outages. Additionally, a mockup of the tube-to-tubesheet joints was fabricated, and indications were machined into the carbon steel tubesheet, inconel cladding, and tube end (beyond the cladding) regions. Figure 1 depicts these regions of the joint design. This mockup and the new definition of the pressure boundary resulted in new, enhanced BWOG ECT Analyst Guidelines providing direction on dispositioning indications
                            .in this region of the tube-to-tubesheet joints. Many indications at ANO-1 and ONS-2 being
                            . identified as SAls in 1998 would have previously been identified as TEAS or MEAS using the 1997
                   .         protocol. ONS-1 and ONS-3 historical data was reviewed using the new ECT analyst guidance.
The licensees identified that these plants had indeed returned tubes to service in the past containing indications within the pressure boundary.
As a result of these developments, a re-analysis of the 1997 CR-3 upper tubesheet rotating coil data has been initiated. The objective is to analyze the CR-3 data using the enhanced ECT analyst guidelines in order to identify any indications which may be reclassified as being within the pressure boundary (Regions 2 and 3 of Figure 1).
The planned sequence of events is as follows:
a ~ Obtain Tuban 11 list of CR-3 TEAS and gather applicable optical disks
                                   =     Obtain mockup used at ONS for ECT evaluation
;
* Develop Analyst Guidelines
                                 ~=      Select Analyst
                                   =     Train Analyst
(                                                                                                                                         l i * : Perform re-analysis of CR-3 ECT data (ongoing)
                                   = Define pressure boundary through analytical techniques
                                   =     Determine if indications are in the pressure boundary
                                       =
             ,                   ia > Perform ' bounding leak ate evaluation and operational assessment l                                                                                                                                         l p
ll l
[h             a       . N..         _     __                      _ _ _ _ , _ _ _ _

U.S. Nuclear Regulatory Commission                                                   Attachment A
      . 3F0698-28                                                                               Page 4 of 8 l         Evaluation '
The following assessment is intended to demonstrate that this issue is not safety significant.
However, this issue may become a compliance issue resulting in shutdown of the reactor.
Therefore, an expedited review of this LAR is warranted to prevent this unnecessary transient.
The root cause of previously incomplete ECT analysis results is lack of having a clearly defined pressure boundary in the tube-to-tubesheet region. This subsequently resulted in the previously used ECT analysis techniques being inappropriate for classifying tube end indications.
FSAR and ITS Information The OTSGs function as part of the reactor coolant system (RCS) pressure inundary. The OTSG tubing provides about 50% of the surface area of the RCS, and as such, stringent design, fabrication, inspection and operational requirements are imposed upon the tubing.
ITS 3.4.12 contains RCS operational leakage limits and surveillance requirements. Normal operational OTSG primary-to-secondary leakage is limited to 150 gallons per day (gpd) through a single OTSG. ITS 3.4.12 Bases record that a conservative value of one gallon per minute (gpm) of primary-to-secondary leakage is assumed for steam line break scenarios.
Final Safety Analysis Report (FSAR) Section 4.3.4 provides an overview of the design basis for the OTSGs. FSAR accident scenarios are described in Sections 14.2.2.1, Steam Line Failure Accident, and 14.2.2.2, Steam Generator Tube Rupture Accident.
Failure Modes With regard to tube burst or rupture considerations, the TEA and MEA indications are contained within regions 1 and 2, as shown in Figure 1. The tubesheet physically limits tube deformation and prevents tube burst or tube rupture. Additionally, the length of the tuly sheet precludes the tubes from being withdrawn in an accident scenario. Thus, these scenarios are not considered credible.
Therefore, the tube structural integrity requirements are satisfied.
The primary failure mode that must be assessed for these indications is accident leakage.
Concurrent with the re-analysis of the ECT data, CR-3 is performing an accident leakage assessment to determine the impact of leaving these indications in service. The operational leakage limit of 150 gpd through a single OTSG is not affected.
The leakage calculation for TEAS'and MEAS will conservatively assume that all known flaws in the most susceptible OTSG will leak at the end of the cycle. The "B" OTSG has 554 tubes with indications in this region, and thus will be modeled as the most susceptible generator. A worst case leakage calculation has been performed. Axial TEA and MEA indications are assigned a leakale value determined from laboratory leakage testing performed at FTI. This leakage value is an average leakage of several mockups which had leakage from machined flaws, and is dependent on

U.S. Nuclear Regulatory Commission                                                     Attachment A
                     . 3FM98-28                                                                                 Page 5 of 8 indication o'rientation. Circumferentially oriented TEA and MEA indications are assigned a leakage value higher than the axial indications. Additionally, half of the circumferential indications l                      are assumed to be located in the outer periphery of the bundle, and are assigned an even higher -
l                      leakage rate to accommodate the affects of tubesheet bow. Preliminary calculations indicate that the cumulative TEA and MEA accident leakage, assuming all 554 tubes leak at the end of the cycle, is between 0.010 and 0.125 gpm.
This leak rate value from the TEAS and MEAS will be added to FPC's previously submitted OTSG operational assessment (FPC to NRC letter,3F0598-08, dated May 18,1998, "An Operational Assessment 01' Steam Generator Tube Degradation at Crystal River Unit 3") leak rate l                      determination. The resultant. cumulative projected end of cycle leakage for the limiting case OTSG, under main steam line break (MSLB) conditions, is between 0.011 and 0.126 gpm.
l lI The FSAR MSLB analysis assumes one gpm leakage in one steam generator as an initial condition.
The dose consequences resulting from the MSLB accident meet the acceptance criteria defined in 10 CFR 100 and bound the potential leakage calculated from leaving the TEAS in service.
Therefore, the high value of 0.126 gpm is bounded in terms of offsite dose.
In conclusion, since the "B" OTSG is bounding, this evaluation demonstrates that the steam generators are capable of performing their intended safety function during normal operation and postulated accident conditions, even with the TEAS and MEAS in service. Additionally, based on the potential leakage from leaving the TEAS and MEAS being bounded by the CR-3 safety analyses, the proposed LAR will not be of potential detriment to the public health and safety.
NO SIGNIFICANT IIAZARDS CONSIDERATION EVALUATION An evaluation of this proposed LAR has been performed in accordance with 10 CFR 50.91(a)(1) regarding significant hazard considerations, using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this LAR follows:
     . s (1) Imvive a sigmficant increase in the probability or consequences of an accident presiously evaluated.
This evaluation addresses the potential effects of operating with TEAS and MEAS within the pressure boundary cladding region. The indications remaining in service are within the upper end of the tube pressure boundary (regions 1 and 2 as shown in Figure 1). Two accidents analyzed in the SAR must be evaluated: Steam Generator Tube Rupture and Main Steam Line Break.
: The steam generator tube rupture accident assumptions bound the possible affects of leaving these indications in service. A complete circumferential severance of a tube is assumed in the accident scenario. The location of these indications in the upper tubesheet precludes a tube rupture from occurring (the tubes are restraincd by the tubesheet). Additionally, in the event of a complete circumferential severance, the tube will not retract from the tubesheet. Thus, t'
C___     __ ______

             ' U.S. Nuclear Regulatory Commission                                                                                                                                   Attachment A
      .      : 3F0698-28                                                                                                                                                                          . Page 6 of 8
                     ' the probabilitiof occurrence of this accident is not increased by leaving these indications in service.
1 The main steam line break accident is not initiated by the condition of the tubing. - However, an assumption of one gpm primary-to-secondary leakage through the OTSG is assumed in the MSLB accident analysis. Calculated cumulative leakage, assuming all of the indications are leaking, is determined to be well' below one gpm,- thus the accident analysis initial assumptions bound the existing condition of the'OTSGs. Thus, it is concluded that the probability of occurrence of a main steam line break is not increased by this change.
                     - Therefore, this change does not . involve a significant increase in the probability or consequences of an accident previously evaluated.
              .'(2)     Create the possibility of a new or dgerent kind of accident from any accident previously emluated.
No new failure modes or accident scenarios are created by allowing operation with TEAS and MEAS extending within the tubes' pressure boundary. The TEAS and MEAS remaining in service are within the upper end of the tube pressure boundary and even in the event of a complete circumferential severance, the tube will not retract from the tubesheet. ; Therefore, the tubesheet hoop effect will still act to minimize leakage. The postulated potential leakage generated from allowing these indications to remain in service is bounded by the CR-3 MSLB scenario. The MSLB scenario has been thoroughly evaluated and the potential damage to the .                                                                                             .f steam generator tubes is not increased. This change does not increase the risk of a plant trip or challenge other safety systems. Therefore, this change does not create a possibility of a new or different kind of accident from any previously evaluated.
(3)      Inwhe a sigmpcant reduction in a margin ofsafety.
ITS Bases 3.4.12 contains relevant information pertaining to the limitations on RCS leakage.
Rese Bases discuss the one gpm primary-to-secondary leakage assumed for a main steam line break accident as well as the steam generator tube rupture accident. As discussed, the maximum calculated accident leakage, assuming all of these indications leak, is well below one gpm. Therefore, the margin of safety as defined in the ITS bases is not significantly
                       , reduced.
l c
ENVIRONLtENTAL IMPACT EVALUATION 10 CFR 51.22(hX9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment I
c to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:-

                                                                                                                                                      )
U.S. Nuclear Regulatory Commission                                                                                     Attachment A
        .    - 3F0698-28                                                                                                                  Page 7 of 8 (i)- ' involve' a significant hazards consideration, -
(ii). result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) result in a significant increase in individual or cumulative occupational radiation '
exposure.
FPC has reviewed this proposed LAR and concludes it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact .
statement or environmental assessment needs to be prepared in connection with this request.
s
                   )
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U.S. Nuclear Regulatory Commission                                                                                             Attachment A
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