ML20059B538

Safety Evaluation Supporting Amend 94 to License NPF-8

ML20059B538
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Site:	Farley
Issue date:	10/20/1993
From:	Office of Nuclear Reactor Regulation
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NUDOCS 9310280277
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.94 TO FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 2 DOCKET N0. 50-364 i

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1.0 INTRODUCTION

By letter dated May 28, 1993, as supplemented on July 29, 1993, September 14, i

1993 and September 22, 1993, Southern Nuclear Operating Company, Inc. (the licensee) submitted a request for changes to the Joseph M. Farley Nuclear i

Plant (Farley), Unit 2, Technical Specifications. The requested amendment revises Technical Specifications 4.4.6.4 and 3.4.7.2, and Bases 3/4.4.6 to allow the implementation of interim steam generator (SG) tube plugging criteria for the tube support plate (TSP) elevations.

The requested amendment reduces the Farley TS limit for specific activity of dose equivalent Iodine 131 as specified in TS 3/4.4.9.

In addition, the amendment reduces the allowed primary-to-secondary operational leakage from any one SG from 500

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gallons per day (gpd) to 150 gpd.

The total allowed primary-to-secondary operation leakage from all three SGs is reduced from one gallon per minute (gpm) or 1440 gpd to 450 gpd. -

The staff reviewed a similar request that was applicable to the ninth operating cycle as documented in Amendment No. 87, dated April 22, 1992 (Reference 1). The staff concluded in Reference 1 that the proposed interim tube repair limits and leakage limits would ensure adequate structural and leakage integrity of the steam generator tubing at Joseph M. Farley Nuclear Plant Unit 2, consistent with applicable regulatory requirements, for the ninth operating cycle.

The following safety evaluation reflects additional information/ operating experience that has been acquired since the staff approved the interim plugging criteria for the ninth operating cycle.

The staff is currently developing a generic interim position on voltage-based limits for outside-diameter stress corrosion cracking at tube support plate elevations.

The staff has recently published several tentative conclusions regarding voltage-based plugging criteria in draft NUREG-1477; however, the staff is continuing to evaluate an acceptable generic position which takes into consideration public comments received on draft NUREG-1477 and additional data which has been made available from European nuclear power plants.

The staff currently plans to document its final position in a generic letter with the associated technical basis being documented in the final version of NUf'.EG-1477.

9310280277 931020 PDR ADOCK 05000364 p

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In the meantime, pending completion and issuance of the staff's generic position on voltage-based interim plugging criteria (IPC), the staff is continuing to evaluate IPC proposals on a case-specific basis, as necessary, j

to ensure that there is adequate assurance of public health and safety.

The staff's current evaluation, documented herein, is consistent with the staff's previous case-specific evaluation of the Farley Unit 2 IPC application with the exception that the staff has requested that the licensee calculate the potential steam generatur tube leakage during a postulated main steam line break (MSLB) in accordance with the methods described in draft NUREG-1477.

2.0 BACKGROUND

The modifications to the tube repair limits, as documented in Reference 1, included a one-volt repair criterion for axially oriented outside-diameter stress corrosion cracking (ODSCC) flaws confined to within the thickness of the tube support plate in lieu of the depth-based limit of 40-percent.

The staff review concluded that the interim tube repair limits and leakage limits would ensure adequate structural and leakage integrity of the steam generator tubing at Joseph M. Farley Nuclear Plant, Unit 2, cor,sistent with applicable regulatory requirements, for the ninth operating cycle.

The licensee's current proposal is applicable to Cycle ten operation and is similar to the licensee's previous proposal which was approved as documented in Reference 1.

3.0 EVALUATION 3.1 Tube Intearity Issues The purpose of the Technical Specification tube repair limits is to ensure that tubes accepted for continued service will retain adequate structural and leakage integrity during normal operating, transient, and postulated accident conditions, consistent with General Design Criteria 14,15, 31 and 32 of 10 CFR Part 50, Appendix A.

Structural integrity refers to maintaining adequate margins against gross failure, rupture, and collapse of the steam generator tubing.

Leakage integrity refers to limiting primary-to-secondary leakage to within acceptable limits. The traditional strategy for accomplishing these objectives has been to establish a minimum wall thickness requirement in accordance with the structural criteria of Regulatory Guide 1.121, " Basis for Plugging Degraded PWR Steam Generator Tubes." Allowance for eddy current measurement error and flaw growth between inspections has been added to the minimum wall thickness requirements (consistent with the Regulatory Guide) to arrive at a depth-based repair limit.

Enforcement of a minimum wall thickness requirement would implicitly serve to ensure leakage integrity (during normal operation and accidents), as well as structural integrity.

It has been recognized, however, that defects, especially cracks, will occasionally grow entirely through-wall and develop small leaks.

For this reason, limits on allowable primary-to-secondary leakage have been established in the Technical Specifications to ensure timely plant shutdown before adequate structural and leakage integrity of the affected tube is impaired.

1he interim tube repair limits for Farley Unit 2 consist of voltage amplitude criteria rather than the traditional depth-based criteria.

Thus, the repair

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criterion represents a departure from the past practice of explicitly enforcing a minimum wall thickness requirement.

The industry-wide data base from the pulled tube examinations show that for bobbin indications at or near 1.0 volt (i.e., the IPC repair limit) maximum crack depths range between 20% and 98% through-wall. The likelihood of through-wall or near through-wall crack penetrations appears to increase with increasing voltage amplitude.

For indications at or near 2.0 volts, the maximum crack depths have been found to generally range between 50% and 100%

through-wall. Clearly, many of the tubes which will be found to.contain non-repairable indications under the proposed interim criteria may develop through-wall and near through-wall crack penetrations during the upcoming cycle, thus creating the potential for leakage during normal operation and postulated MSLB accidents. The staff's evaluation of the proposed repair criteria from a structural and leakage integrity standpoint is provided in i

Sections 3.2 and 3.3, respectively. Section 3.4 contains the staff's evaluation of several inspection issues and Section 3.5 documents the staff's evaluation of the IPC operating experience at Farley Unit 2.

3.2 Structural Inteority

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In support of the 1.0 volt repair limit approved in Reference I for the ninth operating cycle, the licensee developed a burst pressure / bobbin voltage correlation to demonstrate that bobbin indications satisfying the 1.0 volt interim repair criterion would retain adequate structural margins during Cycle 9 operation, consistent with the criteria of Regulatory Guide 1.121. The correlation was developed from both pulled tube data and laboratory tube specimens containing ODSCC flaws. The bobbin voltage data used to construct the burst pressure / bobbin voltage correlation were normalized to be consistent with the calibration standard voltage set-ups and voltage measurement procedures described in WCAP-12871 Revision 2..The normalization was performed to ensure consistency among the voltage data in the burst pressure / bobbin voltage correlation and consistency between the voltage data in the correlation and the field voltage measurements at Farley Unit 2.

For any specific individual tube, voltage measurement uncertainty and/or voltage growth may exceed the value assumed in the previously mentioned Regulatory Guide 1.121 deterministic analysis since the deterministic analysis does not consider the full tails of the voltage measurement uncertainty and voltage growth distributions. Similarly, the burst pressure for some tubes 1

may be less than the 95-percent lower prediction interval values in the burst pressure / bobbin voltage correlation.

These distribution tails may involve I

sizable numbers of tubes in instances where a large number of tubes with

-indications are being accepted for continued service. To directly account for l

these uncertainties, Monte Carlo methods have typically been used to demonstrate that the probability of burst during MSLB accidents is acceptably low for the distribution of voltage indications being left in service. Under this approach, the beginning-of-cycle (B0C) indications left in service are projected to the end-of-cycle (EOC) by randomly sampling the non-destructive examination (NDE) uncertainty probability distribution and the voltage growth per cycle probability distribution.

For each EOC Monte Carlo sample of bobbin voltage, the burst pressure / bobbin voltage correlation is randomly sampled to

4 obtain a burst pressure. A number of Monte Carlo samples (e.g., 100,000) are performed for the entire BOC distribution.

The probability of tube burst at MSLB is obtained as the sum of the samples resulting in burst pressures less than the HSLB pressure differential of 2650 psi divided by the number of times the distribution of indications left in service is sampled.

This kind of Monte Carlo analysis was performed for the distribution of indications found during the 1990 inspection at Farley Unit 2 (prior to implementing the first IPC at Farley).

This analysis indicated that implementation of a 3.6 volt repair criterion at that time would have a conditional probability of burst given a MSLB of approximately 3x10' yielded The staff concurs that this is an extremely low probability, approximately three orders of magnitude less than the value considered in a staff generic risk assessment for steam generators (NUREG-0844).

The staff concludes that the proposed 1.0 volt interim criterion will provide adequate assurance that tubes with indications which are accepted for continued service during Cycle 10 operation will meet the burst pressure criteria of Regulatory Guide 1.121.

The staff notes that the bounding value of voltage growth per cycle at Farley Units 1 and 2 during the 1987 and 1990 time f ram (the most recent outage data was not available) has not exceeded 2.6 volts.

The staff estimates this 2.6 volts to represent a bounding value, assum%g no increase in corrosion rates over what has been observed previously at Farley Units 1 and 2.

Assuming a 20-percent voltage measurement uncertainty (upper 95-percent confidence value previously estimated by the licensee) for a 1.0 volt indication left in service, the E0C voltage is expected by the staff to be bounded by 3.8 volts. This is below the 4.5 volt voltage limit evaluated by the licensee as the lower 95-percent confidence limit for meeting the most limiting burst pressure criterion (i.e., three times normal operating pressure differential) using the most recent burst pressure correlation.

The licensee's current submittal permits bobbin indications greater than 1.0 volt but less than 3.6 volts to remain in service if a motorized rotating pancake coil (MRPC) probe inspection does not detect a flaw, and it requires flaw indications with a bobbin voltage greater than 3.6 volts to be plugged or repaired.

The staff notes that the 3.6 volts reflects an alternate plugging criteria (APC) voltage limit that was derived in WCAP-12871 Revision 2.

Since the issuance of WCAP-12871 Revision 2 in February 1992, additional data has been added to the data base used in the development of this APC voltage limit and several of the existing data points in the data base have been updated as a result of additional analysis. This additional data would result in a lower APC voltage limit for Farley Unit 2.

During the last Unit 2 outage, only 17 tubes had bobbin voltages between 1.0 volt and 3.6 volts with no detectable degradation being observed during the MRPC inspection.

The licensee believes that allowing indications which have no detectable degradation by MRPC inspection and have bobbin voltages between 1.0 volt and 3.6 volts to be a conservative approach for IPC implementation and that revision of the 3.6 volt value due to the additional / revised data is not necessary.

The licensee reached these conclusions, in part, for the following reasons:

1) indications which exhibit no detectable degradation i

5 during MRPC examination will not burst during normal operation since the indication is constrained by a tube support plate, 2) a tube with a bobbin voltage of 3.6 volts would still meet the margin of safety against tube failure under postulated accident conditions contained in the ASME Boiler and Pressure Vessel Code, 3) leakage from tubes exhibiting no detectable degradation by the MRPC will not be significant during a MSLB, and 4) flaws from the most recent Farley Unit I tube pulls with voltages in the range of interest (i.e., approximately 3 volts) that were detectable by both the bobbin coil and the MRPC did not leak at MSLB differential pressures.

The staff notes that short and/or relatively shallow cracks that are detectable by the bobbin coil may sometimes not be detectable by the MRPC probe, although the MRPC probe is considered by the staff to be more sensitive to longer, deeper flaws which are of structural significance.

The staff further notes that burst strength is not a unique function of voltage, rather for a given voltage there is a statistical distribution of possible burst strengths as indicated in the burst pressure / bobbin voltage correlation. The staff believes that the burst pressure for bobbin indications which were not confirmed to be flaw-like by the MRPC probe will tend to be at the upper end of the burst pressure distribution (i.e., exhibit a higher burst pressure).

The 3.6 volt cutoff, such that all bobbin indications would be plugged or repaired (with or without confirming MRPC indications), provides assurance that all excessively degraded tubes will be removed from service.

The staff further notes that the projected leakage from these tubes (i.e., tubes with bobbin voltages between 1.0 and 3.6 volts which exhibited no detectable degradation during the MRPC inspection) will be considered in the leak rate assessment performed by the licensee prior to plant restart.

Thus, the staff finds the proposed exception to the 1.0 volt criterion to be acceptable.

The value for the conditional probability of burst given a MSLB referenced above (i.e., 3x104) was determined from the original burst pressure / bobbin voltage correlation submitted to support the issuance of Reference 1 (i.e.,

WCAP-12871 Revision 2). The staff notes that the licensee will perform an evaluation of the probability of tube burst following the outage, consistent with the requirements in WCAP-12871 Revision 2.

This analysis should' be performed with the most recent burst pressure / bobbin voltage correlation and should consider the most recent growth rate data. The results of this analysis should be submitted to the staff as soon as possible following completion of the refueling outage.

3.3 Leakaae Intearity A number of the indications satisfying the proposed interim 1.0 volt repair limit can be expected to have, or to develop, through-wall and/or near through-wall crack penetrations during the next cycle, thus creating the potential for primary-to-secondary leakage during normal operation, transients, or postulated accidents.

The staff finds that adequate leakage integrity during normal operating conditions is assured by the proposed Technical Specification limits on allowable primary-to-secondary leakage.

Adequate leakage integrity during transients and postulated accidents is demonstrated by showing that for the most limiting accident, assumed to occur

6 at the end of the next cycle, the resulting leakage will not exceed a rate that will result in offsite dose limits being exceeded.

As the basis for estimating the potential leakage during MSLB accidents, Westinghouse has correlated leakage test data obtained under simulated MSLB conditions with the corresponding bobbin voltage amplitudes.

The correlation is based on a linear regression fit of the logarithms of the corresponding leak rates and bobbin voltages. The leak rate data exhibits considerable scatter relative to the mean correlation.

Thus, prediction intervals for leak rate at a given voltage have been established to statistically define the range of potential leak rates.

As part of the on-going review of the valtage-based repair criterion for ODSCC, the staff is continuing to review the correlation of the leak rate data to bobbin voltage. The staff has tentatively concluded that no proven relationship between leakage rate and voltage presently exists and that the proposed approach fails to account for non-detected ODSCC that remains in service.

However, until the issue of the leak rate versus voltage correlation is resolved, the staff has concluded that a voltage-based approach can be used if these non-conservatisms are accounted for and sufficient conservatisms are included in the analysis.

Therefore, at i

the staff's request, the licensee has committed to provide a calculation of potential MSLB leakage by a methodology designed to address the staff concerns. The methodology that the licensee will use to calculate the MSLB leakage is described in draft NUREG-1477.

The staff notes that the MSLB leakage analysis should be performed with the most recent leak rate data for 7/8-inch outside diameter tubing.

In addition, the voltage growth distribution used in the leakage assessment should 1) consider the most recent voltage growth data (i.e., Cycle 9) and 2) be adjusted for the planned Cycle 10 duration.

3.4 Inspection Issues In support of the proposed interim repair limit, the licensee proposes to utilize the edd) nurrent test guidelines provided in Attachment 5 to the licensee's July 29, 1993, submittal to ensure the field bobbin indication voltage measurements are obtained in a mannur consistent with how the voltage limit was developed.

These guidelines define, in part, the bobbin specifications, calibration requirements, specific acquisition and m alyses criteria, and flaw recording guioelines to be used for the inspection of the steam generators. Attachment 5 contains several enhancements / modifications to the guidelines proposed in WCAP-12871 Revision 2 including, in part, requirements to 1) record all indications regardless of voltage amplitude, 2) perform MRPC inspections of 100 tubes, including all tubes with dent indications exceeding 5 volts as measured by the bobbin coil and also including tube support plate intersections with artifact indications or indications with unusual phase angles (expansion of this sample, if required, will be based on the nature and number of the flaws discovered), 3) perform MRPC examinations of all tubes with bobbin voltages in excess of 1.0 volt, and

4) inform the staff prior to Cycle 10 operation of any unexpected MRPC findings relative to the assumed characteristics of the flaws at the tube support plates (which includes any detectable circumferential indications or detectable indications extending outside the thickness of the tube support

7 plate) and provide a safety evaluation, if applicable, to address these findings.

The licensee's submittal primarily references the eddy current analyst guidelines in Appendix A to WCAP-12871 Revision 2.

Since the issuance of WCAP-12871, several modifications have been made to these guidelines as documented in WCAP-12985 Revision 2, "Kewaunee Steam Generator Tube Plugging Criteria for ODSCC at Tube Support Plates." The staff believes that several of the modifications made in the most recent eddy current analyst guidelines may enhance the inspection program at Farley and should be considered for inclusion in future outages. The staff notes, however, that the original calibration procedure for the bobbin coil in WCAP-12871 Revision 2, which requires setting the bobbin coil voltage amplitude from the 400/100 kHz differential channel from the four 100% through-wall holes, is preferred over the more recent guidelines which require calibration on the four 20% through-wall holes, as discussed in draft NUREG-1477.

Furthermore, the staff notes that the calibration procedure used to analyze the field eddy current data during the E0C 9 refueling outage should be consistent with the calibration procedure used in the development of the burst pressure / bobbin voltage correlation.

3.5 Assessment of IPC Methodoloav The staff notes that the methodology described in Reference 1 and in this safety evaluation for predicting MSLB leakage and the probability of rupture given a MSLB depends largely, in part, on the ability to accurately predict an i

E0C voltage distribution. An assessment of the effectiveness of the

'l methodology described in WCAP-12871 Revision 2 and in this safety evaluation

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for predicting the E0C voltage distribution is warranted to further ensure the adequacy of the methodology used. The assessment for Farley Unit 2 should be j

made in a manner consistent with the methodology described in WCAP-12871 Revision 2 and in this safety evaluation.

The assessment should address any discrepancies between the predicted and actual values.

The following information should be included in this assessment in both tabular and graphical form:

1 E0C 8 voltage distribution - all indications found during the a.

inspection regardless of MRPC confirmation b.

Cycle 8 growth rate (i.e., from BOC 8 to E0C 8) c.

E0C 8 repaired indications voltage distribution - distribution of indications presented in (a) above that were repaired (i.e., plugged or sleeved) d.

Voltage distribution for indications.left in service at the B0C 9 regardless of MRPC confirmation - obtained from (a) and (c) above e.

Voltage distribution for indications left in service at the BOC 9 that were confirmed by MRPC to be crack-like or not MRPC inspected f.

Non-destructive examination uncertainty distribution used in predicting the E0C 9 voltage distribution g.

Projected E0C 9 voltage distribution using the methodology in WCAP-12871 Revision 2 h.

Actual E0C 9 voltage distribution - all indications found during the inspection regardless of MRPC confirmation

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Cycle 9 growth rate (i.e., from B0C 9 to E0C 9) j.

E0C 9 repaired indications voltage distribution - distribution of indications presented in (h) above that were repaired (i.e., plugged or sleeved) k.

Voltage distribution for indications left in service at the BOC 10 regardless of MRPC confirmation - obtained from (h) and (j) above 1.

Voltage distribution for indications left in service at the BOC 10 j

that were confirmed by MRPC to be crack-like or not MRPC inspected Non-destructive examination uncertainty distribution used in m.

predicting the E0C 10 voltage distribution Projected E0C 10 voltage oistribution using the methodology in WCAP-n.

12871 Revision 2 The staff recognizes that compilation of the assessment on the overall IPC methodology may not be possible until after completion of the refueling outage.

Currently available information (e.g., items 3a to 3g inclusive) should be submitted to the NRC as soon as possible along with a proposed schedule of when a complete assessment could be provided to the NRC staff.

3.6 Radioloaical Consecuences As part of the Farley IPC TS request, SNC proposed that allowable limits for specific activity of reactor coolant contained in Technical Specifications be 1

reduced by a factor of four to enable a factor of four increase in allowable post-MSLB primary-to-secondary leakage.

SNC concluded that the increased leakage estimates would be offset by the reduced Technical Specification limits on allowable reactor coolant activity. By letter dated September 22, 1993, the licensee submitted a revised TS page 3/4 4-26 in response to a staff comment that the Figure on this page should reflect a constant ratio of allowable iodine concentration to power over the power operating range.

The base analysis for the SNC proposal was provided in SNC's June 4,1992, letter (SNC response to May 20, 1992 staff request for additional-information).

This analysis determined the maximum permissible steam generator (SG) primary-to-secondary leak rate during a main steam line break (MSLB) for both Farley units considering both the pre-accident and event-generated iodine spike cases. The licensee, in performing its analyses, considered the acceptance criteria of SRP Sections 15.1.5 Appendix A. As a i

result of the June 4,1992 analysis, the licensee concluded that the limiting

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primary-to-secondary SLB leakage would be governed by the event-generated 1

spike case and should be limited to 5.7 gpm so that accident consequences remain within SRP acceptance criteria.

The present request reduces the allowable limits for reactor coolant system specific activity by a factor of four, in order to allow an increase in SG

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1eakage during a postulated MSLB (calculated per section 3.3 above) by a 4

similar factor of four (above the June 4,1992 leakage limit of 5.7 gpm) and still meet SRP limits.

The staff concludes that no increased radiological j

consequences would result from the increased projected leakage since the allowable Technical Specifications specific activity limits are being reduced accordingly. The calculated MSLB leakage as determined by the methodology discussed in section 3.3. of this safety evaluation must be below the proposed

9 leakage limit (or additional tubes must be repaired until the leakage is within limits).

Based on the above, we find the proposed changes acceptable.

4.0 EXIGENT CIRCUMSTANCES

The Commission's regulations,10 CFR 50.91, contain provisions for issuance of amendments when the usual 30-day public notice period cannot be met. One type of special exception is an exigency. An exigency is a case where the staff and licensee need to act promptly and time does not permit the Commission to publish a Federal Register notice allowing 30 days for prior public comment, and it also determines that the amendment involves no significant hazards considerations.

The staff has determined that exigent circumstances exist in that a failure to act promptly would result in a delay in startup from the current Farley Unit 2 refueling outage. The requested TS amendments are required to support performance of the surveillances necessary to declare the Farley Unit 2 steam j

generators operable prior to proceeding to mode 4.

The licensee explained this and why the exigency cannot be avoided. The staff determined that the licensee's TS application was timely and that the licensee did not create the exigency by reason _of the time at which it filed the application.

In addition, there was no indication that the licensee failed to use its best efforts to make a timely application in order to create the exigency and to take advantage of the procedures outlined in 10 CFR 50.91(a)(6).

The Commission notified the public by publishing a notice in the Federal Register on October 5, 1993 (58 FR 51889). The notice provided an opportunity for hearing and allowed 15 days for public comments on a proposed determination of no significant hazards consideration. There were no public comments in response to the notice published in the Federal Register.

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The amendment changes Technical Specifications 4.4.6.4 and 3.4.7.2, and Bases 3/4.4.6, to allow the implementation of interim steam generator tube plugging criteria for the tube support plate elevations.

The amendment reduces the Farley TS limit for specific activity of dose equivalent Iodine 131 as specified in TS 3/4.4.9.

In addition, the amendment reduces the allowed primary-to-secondary operational leakage from any one steam generator from 500 gallons per day to 150 gallons per day.

The total allowed primary-to-secondary operational leakage through all steam generators is reduced from one gallon per minute (1440 gallons per day) to 450 gallons per day.

This amendment is only applicable for the tenth Farley Unit 2 operating cycle.

The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility in accordance with the amendment would not:

(1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or

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10 (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The licensee has provided the following analysis regarding no significant hazards considerations using the Commission's standards.

(1) Operation of Farley Unit 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

1 Testing of model boiler specimens for free standing tubes at room temperature conditions show burst pressures as high as approximately 5000 psi for indications of outer diameter stress corrosion cracking with voltage measurements as high as 26.5 volts.

Burst testing i

performed on pulled tubes with up to 10 volt indications show burst pressures in excess of 5900 psi at room temperature.

Correcting for the effects of temperature on material properties and minimum strength levels (as the burst testing was done at room temperature),

tube burst capability significantly exceeds the [ Regulatory Guide)

R.G. 1.21 criterion requiring the maintenance of a margin of three times normal operating pressure differential on tube burst if through-wall cracks are present.

Based on the existing data base, this criterion is satisfied with bobbin coil indications with signal amplitudes several times the 1.0 volt interim plugging criteria, regardless of the indicated depth measurement. This structural limit is based on a lower 95% confidence level limit of the data, i

The 1.0 threshold volt criteria provides an extremely conservative margin of safety to the structural limit considering expected growth rates of ODSCC [outside diameter stress corrosion cracking) at Farley. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by a burst pressure to voltage correlation.

However, relative to expected leakage during 4

normal operating conditions, no field leakage has been reported from tubes with indications with a voltage level of under 7.7 volts for a 3/4 inch tube with a 1.0 volt correlation to 7/8 inch tubing (as compared to the 1.0 volt proposed interim tube plugging limit).

Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary-to-secondary leakage and steam release to the environment are Loss of External Electrical Loan and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Sdcondary System Pipe Failure, Steam Generator i

Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism Housing.

Of these, the Major Secondary System Pipe Failure is the most limiting for Farley in considering the potential for off-site doses. The offsite dose analyses for the other events which model primary-to-secondary leakage and steam release from the secondary side to the environment assume that the secondary side remains intact.

The steam generator tubes are not subjected to a sustained increase in differential pressure, as is

11 the case following a steam line break event.

This increase in differential pressure is responsible for the postulated increase in leakage and associated offsite doses following a steam like break event. Upon implementation of the interim plugging criteria, it I

must be verified that the expected distribution of cracking indications at the tube support plate intersections are such that i

primary-to-secondary leakage would result in site boundary dose l

within the current licensing basis.

Data indicate that a threshold voltage of 2.8 volts would result in through-wall cracks long enough to leak at SLB [ steam line break] conditions.

Application of the proposed plugging criteria requires that the current distribution of a number of indications versus voltage be obtained during the refueling outages.

The current voltage is then combined with the rate of change in voltage measurement to establish an end of cycle voltage distribution and, thus, leak rate during SLB pressure differential.

If it is found that the potential SLB leakage for degraded intersections planned to be left in service is greater than the current licensing basis limit, then additional tubes will be plugged or repaired to reduce SLB leakage potential to within the acceptance limit.

(2)

The proposed license' amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Implementation of the proposed interim tube support plate elevation

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steam generator tube plugging criteria does not introduce any significant changes to the plant design basis.

Use of the criteria does not provide mechanism which could result in an accident outside of the region of the tube support plate elevations.

Neither a single or multiple tube rupture event would be expected in a steam generator in which the plugging criteria has been applied (during all plant conditions).

The bobbin probe signal amplitude plugging criteria is established such that operational leakage or excessive leakage during a postulated steam line break condition is not anticipated.

SNC [ Southern Nuclear Operating Company) will implement a maximum leakage rate limit of 150 gpd per steam generator on Unit 2 to help preclude the potential for excessive 1eakage during all plant conditions upon application of the plugging i

criteria.

The R.G.1.121 criterion for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture.

The 150 gpd limit provides for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest permissible crack length.

R.G. 1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded.

The longest permissible crack is the length that provides a factor of safety of three against bursting at normal operating pressure differential. A voltage amplitude of

12 approximately 4 volts for typical ODSCC corresponds to meeting this tube burst requirement at the 95% prediction interval on the burst correlation. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by the burst pressure versus voltage correlation.

Consequently, typical burst pressure versus through-wall crack length correlations are used below to define the " longest permissible crack" for evaluating operating leakage limits.

The single through-wall crack lengths that result-in _ tube burst at three times normal operating pressure differential and SLB conditions are about 0.42 inch and 0.84 inch, respectively.

Normal leakage for these crack lengths would range from 0.11 gallons per minute to 4.5 gallons per minute, respectively, while lower 95%

confidence level leak rates would range from about 0.02 gallons per minute to 0.02 gallons per minute to 0.6 gallons per minute, respectively.

An operating leak rate of 150 gpd will be implemented in application of the tube plugging limit.

This leakage limit provides for detection of 0.4 inch long cracks at nominal leak rates and 0.6 inen long cracks at the lower 95% confidence level leak rates. Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for SLB conditions at leak rates less than a lower 95%

confidence level and for three times normal operating pressure differential at less than nominal leak rates.

(3)

The proposed license amendment does not involve a significant reduction in margin of safety.

The use of the interim tube support plate elevation plugging criteria is demonstrated to maintain steam generator tube integrity commensurate with the requirements of R.G. 1.121.

R.G. 1.21 describes a method acceptable to the NRC staff for meeting GDCs (General Design Criteria] 2, 14, 15, 31, and 32 by reducing the probability of the consequences of steam generator tube rupture.

This is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established.by inservice inspection, for which tubes with unacceptable cracking should be removed from service: Upon implementation of the criteria, even under the worst case conditions, the occurrence of ODSCC at the tube support plate elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions. The most limiting effect would be a possible increase in leakage during a steam line break event.

Excessive leakage during a steam line break event, however, is precluded by verifying that, once the criteria are applied, the expected end of cycle distribution of crack indications at the tube support plate elevations would result in minimal, and acceptable primary to secondary leakage during the event, and hence, help to demonstrate

13 radiological conditions are less than a small fraction of the 10 CFR 100 guideline,

)

i In addressing the combined effects of LOCA + SSE [ loss-of-coolant accident plus safe shutdown earthquake] on the steam generator component (as required by GDC 2), it has been determined that tube collapse may occur in the steam generators at some plants.

This is the case as the tube support plates may become deformed as a result of lateral loads at the wedge supports at the periphery of the plate due to either the LOCA rarefaction wave and/or SSE loadings.

Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.

Additionally, the margin to burst for the tubes using the interim plugging criteria is comparable to that currently providing by existing technical specifications.

There are two issues associated with steam generator tube collapse.

First, the collapse of steam generator tubing reduces the RCS flow area through the tubes.

The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in turn, may potentially increase Peak Clad Temperature (PCT).

Second, there is a potential the partial through-wall cracks in tubes could i

progress to through-wall cracks during tube deformation or collapse.

Consequently, a detailed leak-before-break analysis was performed and it was concluded that the leak-before-break methodology (as permitted by GDC4) is applicable to the Farley Unit I and 2 reactor i

coolant system primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design basis of the plant.

Excluding breaks in the RCS primary loops, the LOCA loads from the large branch line breaks were analyzed at Farley Unit I and 2 and were found to be of insufficient magnitude to result in steam generator tube collapse or significant deformation.

Regardless of whether or not leak-before-break is applied to the primary loop piping at Farley, any flow area reduction is expected to be minimal (much less than 1%) and PCT margin is available to account for this potential effect.

Based on analyses results, no tubes near wedge locations are expected to collapse or deform to the i

degree that secondary to primary in-leakage would be increased over current expected levels.

For all other steam generator tubes, the possibility of secondary-to-primary leakage in the event of a LOCA +

SSE event is not significant.

In actuality, the amount of secondary-to-primary leakage in the event of a LOCA + SSE is expected to be less than th&t currently allowed, i.e., 500 gpd per steam generator.

Furthermore, secondary-to-primary in-leakge would be less than primary-to-secondary leakage for the same pressure differential since the cracks would tend to tighten under a secondary-to-primary pressure differential. Also, the presence of l

t 14 the tube support plate is expected to reduce the amount of in-leakage.

Addressing the R.G. 1.83 ansiderations, implementation of the tube plugging criteria is. supplemented by 100% inspection requirements at the tube support plate elevations having ODSCC indications, reduced operating leak rate limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotating pancake coil inspection requirements for the larger indications left in service to characterize the principal degradation mechanism as ODSCC.

As noted previously, implementation of the tube support plate elevation plugging criteria will decrease the number of tubes which must be taken out of service with tube plugs or repaired.

The installation of steam generator tube plugs or tube sleeves would reduce the RCS flow margin, thus implementation of the interim plugging criteria will maintain the margin of flow that would otherwise be reduced through increased tube plugging or sleeving.

Based on the preceding analysis, the licensee determined that the proposed change to the TS would not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any previously evaluated or involve a i

significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis.

Accordingly, the staff finds that the requested amendments do not involve a significant hazards consideration.

6.0 SUNMARY Based on the above evaluation, the staff concludes that adequate structural integrity of the steam generator tubing is ensured for Cycle 10 at Farley Unit 2, consistent with applicable regulatory requirements.

In addition, the staff concludes that the methodology for determining the expected primary-to-secondary leakage during a postulated MSLB at the end of fuel Cycle 10 for Farley Unit 2 is acceptable.

The staff's approval of the proposed interim repair limit is based on the licensee being able to demonstrate that the primary-to-secondary leakage during a postulated MSLB will be acceptable.

The licensee has agreed to report, prior to restart from the ninth refueling outage, the results of the MSLB leakage analysis.

The licensee has also agreed to inform the NRC prior to plant restart from the refueling outage of any unexpected inspection findings relative to the assumed characteristics of the flaws at the tube support plates.

This includes any detectable circumferential indications or detectable indications outside the tube support plate thickness.

The staff requests that any unexpected inspection findings during this outage (e.g., actual F0C voltage distribution differing from predicted E0C voltage distribution) should be discussed with the NRC prior to plant restart.

J A

J 15 7.0 REFERENCE 1.

Amendment No. 87 to Facility Operating License No. NPF-8 Regarding Steam Generator Tube Interim Plugging Criteria for Joseph M. Farley Nuclear Plant, Unit 2, and a correction letter for the amendment package dated April 22, 1992.

8.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of Alabama official was notified of the proposed issuance of the amendment. The State official had no comments.

9.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (58 FR 51889).

Accordingly, the amendaent meets the eligibility criteria for categorical exclusion set forth in 10 CFR i

Sl.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

10.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: K. Karwoski K. Eccleston T. Reed Date:

October 20, 1993 i

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