3F0698-25, Advises That FPC May Be Requesting Enforcement Discretion or Other Regulatory Action in Near Future,Based on Review of Industry Operating Experience.Evaluation of Crystal River, Unit 3,operation W/Tube End Anomalies,Encl

Advises That FPC May Be Requesting Enforcement Discretion or Other Regulatory Action in Near Future,Based on Review of Industry Operating Experience.Evaluation of Crystal River, Unit 3,operation W/Tube End Anomalies,Encl

ML20249B654
Person / Time
Site:	Crystal River
Issue date:	06/16/1998
From:	Cowan J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0698-25, 3F698-25, NUDOCS 9806240022
Download: ML20249B654 (12)

Text

__ _ _

l Fl2rida '

Power CORPORATION

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o m .. ww June 16,1998 3F0698-25 U.S. Nuclear Regulatory Commission Atta: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 Review of Industry Operating Experience Regarding Tube End Anomalies

Dear Sir:

The purpose of this submittal is to inform the NRC that, based on an in-progress review of recent industry operating experience, Crystal River Unit 3 (CR-3) may be requesting enforcement discretion or other regulatory action in the near future. Recent industry experience from Arkansas Nuclear One (ANO) and Oconee Nuclear Station (ONS) revealed that steam generator tube eddy current testing (ECT) indications previously identified as Tube End Anomalies (TEAS) had the potential to extend into the cladding area of the tubing, and thus the pressure boundary. Previous industry guidance was to treat these suspect indication TEAS as acceptable as-is. This industry guidance is under review based on the recent operating experience.

in response to this recent industry information, a re-analysis of the CR-3 ECT data from the 1997 Once Through Steam Generator (OTSG) inservice inspection is in progress at Framatome Technologies, Inc. Final confirmation of the tube ECT data and location of the indications with I respect to the pressure boundary is not expected to be complete until later this week or early next l week. Ilowever, Florida Power Corporation (FPC) recognizes that based on a worst case analysis l of preliminary data, 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.

Based on this patential condition of noncompliance, assuming wo'.st case analysis, FPC has J performed an evaluation of the safety significance and potential consequences of operating with these indications within the steam generator tube pressure boundary. 'B.e Attachment to this letter s o

l contains CR-3's safety analyses for concluding that the operation with the indications found in the go '

l initial data review will not be of detriment to the public heahh and safety, and that neither an l unreviewed safety question nor a sig..ificant hazard consideration may be involved. This evaluation L is being submitted at this time for your information and review. Future actions such as a request l for enforcement discretion or other regulatory action will be determined after the complete review of ECT data as indicated above. FPC will continue to apprise the NRC staff on the results of the data re-analysis and any schedule revisions.

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! 9806240022'980616 DR ADOCK 05000302 3syso w, po , un. sereet ~ cryit.: River, FL s442s470s (3s2) 7ss44ss PDR A riorssa progress company L . - _ _ _ _ . _ _ _ . _ _ _ _ _

, U.S. Nuclear Regulatory Commission Page 2 of 2 If you have any questions regarding this submittal, please contact Ms. Sherry Bernhoft, Manager, Nuclear Licensing at (352) 563-4566.

Sincerely, pflbe - m John Paul Cowan Vice President Nuclear Operations

' JPC/lve

' Attachment xc: Regional Administrator, Region 11 Senior Resident Inspector NRR Project Manager

4 U.S. Nuclear Regulatory Commission Attachment 3F0698-25 Page 1 of 9 A'ITACIIMENT CRYSTAL RIVER UNIT 3 (CR-3)

EVALUATION OF OPERATION WITII TUBE END ANOMALIES NRC Administrative letter 95-05 has been used to develop this information.

1. Tecimical Specification that nill be violated:

The Technical Specification which could potentially be violated is Improved Technical Specification (ITS) 5.6.2.10.4.b which states:

"The OTSG shall be determined OPERABLE after completing the corresponding actions (plug or sleeve all tubes exceeding the plugging / sleeving limit and all tubes mntaining thmugh-wall cracks)... "

Plugging / Sleeving Limit is defined by ITS 5.6.2.10.4.a.8 as:

" Plugging / Sleeving Limit means the extent of degradation beyond which the tube shall be mstored to serviceability by the installation of a sleeve or rtmoved fmm service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube or sleeve wall thickness..."

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 0:once Nuclear Station (ONS). This operating experience indicated the potential for some indications previously identified as Tube End Anomalies (TEAS) and Multiple Tube End Anomalies (MEAS) in the upper roll expansions to extend into the pressure boundary of the tube. The details of this investigation are described in Item 2 below.

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2. Circumstances surmunding the situation including root causes, the needfor prompt action and identification of any relemnt historical events.

Based on information provided by ANO and ONS, Crystal River Unit 3 (CR-3) initiated l Precursor Card 3C-98-2857 on June 9,1998, to document industry operating experience j

pertaining to in-service steam generator eddy current indications which may exceed the improved Technical Specifications (ITS) plugging limit. The indications in question are referred to as Tube End Anomalies (TEAS) and Multiple Tube End Anomalies (MEAS).

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4 U.S. Nuclear Regulatory Commission Attachment 3F0698-25 Page 2 of 9 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 I historically been used only on indications which have been considered to be located outside l the pressurc boundary, and thus the indications could be left in service. If the indications l 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.

l This evaluation addresses the potential that these indications referred to as TEAS and MEAS j may indeed be in the pressure boundary, and, if so, the safety signifance of this potential.

Background

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 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 what defined the pressure boundary,273 tubes in the "A" OTSG and 554 tubes in the "B" OTSG were returned 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, FTI 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 roll expansion regions more closely in spring 1998 outages.

Additionally, a mockup of the tube-to-tubesheet joints was fabriceted, 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 definitian of the pressure boundary resulted in new, enhanced BWOG ECT Analysis Guidelines providing direction on disp sitioning 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 analysis guidance. The licensees

4 U.S. Nuclear Regulatory Commission Attachment 3F0698-25 Page 3 of 9 identified that these plants had indeed returned plants to service in the past containing indications in 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 l ECT analyst guidelines in order to identify any indications which may extend into the pressure boundary (Regions 2 and 3 of Figure 1).

The planned sequence of events is as follows:

= Obtain Tuban Il list of CR-3 TEAS and gather applicable optical disks

= Obtain mockup used at ONS for ECT evaluation a Develop Analyst Guidelines

= Select Analyst

= Trained Analyst

= Perform re-analysis of CR-3 ECT data (ongoing)

= Denne pressure boundary through analytical techniques

= Determine if indications are in the pressure boundary

= Perform a bounding leak rate evaluation and operational assessment The following assessment is intended to demonstrate that this issue is not safety significant.

Ilowever, this issue may become a compliance issue resulting in shutdown of the reactor.

Therefore, an expedited review is warranted to prevent this unnecessary transient.

The root cause 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.

3. Safety bases of the request, including an evaluation of the safety sigmficance and potential consequences of the proposed course of action. This evaluation should include at least a qualitative risk assessment derivedfrom the Licensee's PRA.

FSAR and ITS Infonnation The OTSGs function as part of the reactor coolant system (RCS) pressure boundary. 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.

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U.S. Nuclear Regulatory Commission Attachment 3F0698-25 Page 4 of 9 ITS 3.4.12 contains RCS operational leakage limits and surveillance requirements. Nonnal 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 areas 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 tubesheet 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 deterministic leakage calculation is being performed. Axial TEA and MEA indications are assigned a leakage 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 indication orientation. Circumferentially l oriented TEA and MEA indications are assigned a leakage value higher than the axial l indications. Additionally, half of the circumferential indications are assumed to be located in the outer periphery of the bundle, and are assigned an even higher 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.

l l This leak rate value from the TEAS and MEAS will be added to FPC's previously submitted j OTSG operational assessment leak rate determination. The resultant cumulative projected j end of cycle leakage for the limiting case OTSG, under main steam line break (MSLB) conditions, is between 0.011 and 0.126 GPM.

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U.S. Nuclear Regulatory Commission Atachment 3F0698-25 Page 5 of 9 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 analysis will demonstrate 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.

4. 1he basisfor FPC's conclusion that the noncompliance uill not bc ofpatential detriment to the public health and safety and that neither an unreviewed safety question nor a sigmficant hazard consideration is inwived.

Unreviewed Safety Question and No Significant llazards Review:

(1) Could the proposed activity increase the probability of occurrence of an accident evaluated in the SAR?

No. 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 (areas 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 rupturc from occurring (the tubes are restrained by the tubesheet). Additionally, in the event of a complete circumferential severance, the tube will not retract from the tubesheet. Thus, the probability of occurrence of this accident is not increased by leaving tnese indications in service.

The main steam line break accident is not initiated by the condition of the tubing.

Ilowever, an assumption of one gpm primary-to-secondary leakage through the OTSG is assumed in the 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 ;me break is not increased by this change.

U.S. Nuclear Regulatory Commission Attachment 3F0698-25 Page 6 of 9 (2) Could the proposed activity increase the consequences of an accident previously evaluated in the SAR?

No. 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 (areas 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 generaer tube rupture accident assumptions bound the possible affects of leaving these indicawns 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 restrained by the tubesheet). Additionally, in the event of a complete circumferential severance, the tube will not retract from the tubesheet. Thus, the consequences of this accident already bound the possible consequences by leaving these tubes in service.

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 accident analysis. Calculated cumulative leakage, asswning 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 consequences of a main steam line break are not increased by this change.

(3) Could the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR?

No. Operating with a number of potential TEAS extending within the tubes' pressure boundary has been demonstrated to have a minimal effect on leakage due to the hoop effect of the tubesheet (tube constraint) which also prevents any occurrence of tube burst. The postulated leakage is based on previous leak tests conducted for ANO. The engineering evaluation concluded the OTSGs can perform their pressure boundary safety function and the potential leakage is significantly less than postulated for the CR-3 MSLB scenario. This activity does not change any plant systems, structures or components (SSCs). No valve manipulations, alignments or system configurations have to be altered. Therefore, this activity does not increase the probability of malfunction of equipment important to safety.

U.S. Nuclear Regulatory Commission Attachment 3F0693-25 Page 7 of 9 (4) Could the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in die SAR?

No. The activity will not adversely affect equipment necessary to place the plant in safe shutdown or to mitigate the consequences of an accident. While the TEAS are within the upper pressure boundary portion of the tubes, they do not compromise the tube integrity because the tubes are restrained from bursting due to the hoop effects of the tubesheet. leakage due to operating with a number of tubes with TEA pluggable indications is well below, and bounded by, the one gpm assumption for the CR-3 MSLB scenario. Therefore, this activity does not increase the consequences of malfunction of equipment important to safety.

(5) Could the proposed activity create the possibility of an accident of a different type than any previously evaluated in the FSAR?

No. No new failure modes or accident scenarios are created by allowing operation with TEAS extending within the tubes' pressure boundary. The TEAS 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 TEAS 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 steam generator tubes is not increased. This activity does not increase the risk of a plant trip or challenge other safety systems.

(6) Could the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the SAR?

No. Operation with the indications in areas 1 and 2 (as shown on Figure 1) has a minimal affect on tube integrity. Burst is precluded by the physical restraint of the tubesheet, and leakage has been calculated to be minimal. There is no physical challenge to the plant SSCs by leaving these indications in service. There are no required changes to operating procedures. No new equipment or components were installed. No credible new faiLires are introduced. Therefore, this activity does not create the possibility of a different type of malfuncuon. .

(7) Could the proposed activity reduce tl.e margin of safety as defined in the bases for any improved Technical Specification?

No. ITS Bases 3.4.12 contains relevant information pertaining to the limitations on RCS leakage. These Bases discuss the one gpm primary-to-secondary leakage assumed

i U.S. Nuclear Regulatory Commission Attachment 3F0698-25 Page 8 of 9 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 de6ned in the ITS bases has not been reduced as a result of this activity.

Based on the preceding evaluation, the noncompliance with ITS 5.6.2.10.4.b will not be of potential detriment to the public health and safety and neither an unreviewed safety question nor a significant hazards consideration is involved.

5. The basisfor FPC's conclusion that noncompliance will not involve adverse consequences to the environment.

10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:

(i) involve a significant hazards consideration, (ii) result in a signi6 cant change in the types or signi6 cant 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 potential regt.~t 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.

6. A statement that the request has been approved by the Plant Resiew Committee.

On June 15, 1998, the Plant Review Committee reviewed and approved the safety signi6cance assessment in item 4 above.

7. Any other infonnation the NRC staf deems necessary before making a decision to exercise enforcement discretion.

On February 2 and 3,1998, as CR-3 was increasing in power after an extended outage, RM-G26 (the N-16 radiation monitor for the main steam line from the "B" OTSG) spiked up when reactor power was increased. On February 5,1998, after a main turbine trip l

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U.S. Nuclear Regulatory Commission Attachment 3F0698-25 Page 9 of 9 (approximately at 20% power), another N-16 spike was detected by RM-G26. A slight pressure transient later that day caused a second N-16 spike. Subsequent power changes have not shown correlating N-16 spikes. In two of these events, a confirmation of this leakage was received when RM-Al2 (main condenser vacuum pump exhaust radiation monitor) also trended upwards, indicating an increase in radioactive non-condensable gases in the main condenser. These perturbations were tracked under the CR-3 Corrective Action Program.

Subsequently, there have been other N-16 spikes but their frequency is now approximately at two week intervals. The highest spike observed to date on RM-Al2 has been calculated to be 55 gpd (well below the ITS limit of 150 gpd). FPC has evaluated potential sources for the leakage. The evaluation considered these possible leakage paths:

a) tube plugs with corresponding through-wall defects in tubes previously removed from service b) damaged tube-to-tubesheet joints c) tubes with leaking through-wall defects d) a ruptured or severed tube c) leaking sleeve (s)

FPC considers that either previously damaged plugs (both rolled and welded) or damaged tube-to tubesheet joints are the most probable sources for OTSG primary-to-secondary leakage. There is site specific operating history to support the conclusion that indicated primary-to-secondary leakage will stabilize at low value, most probably less than 20 gpd (latest leakage data shows 14 days of operation with approximately 3 gpd). Based upon the low levels of primary-to-secondary leakage and the low probability of identifying the sources, CR-3 determined to continue power operations with a heightened awareness on monitoring primary-to-secondary leakage. This heightened awarenes; will continue for the remainder of the operating cycle. FPC will repair or plug affected tubes during Refueling Outage 11.

Additionally, fuel cladding integrity has been well maintained this cycle based on Dose 3

Equivaler.t (DE) 1-131 levels remaining stable below 5 x 10 pCi/cc. This is well below ITS LCO 3.4.15.

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