ML20211D844

SER Approving SG 90-day Rept Submitted by Duquesne Light Co for Beaver Valley Power Station,Unit 1,per GL 95-05

ML20211D844
Person / Time
Site:	Beaver Valley
Issue date:	09/17/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20211D843	List: ML20211D839 ML20211D844
References
GL-95-05, GL-95-5, NUDOCS 9709290258
Download: ML20211D844 (5)

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REVIEW BY THE OFFICE OF NUCLEAR REACTOR REGULATION I RELATED TO STEAM GENERATOR 90-DAY REPORT SUBMITTED BY PUOVESNE llGHT COMPANY BEAVER VALLEY POWER STATION. UNIT NO. 1 DOCKET NO. 50-334 1.0 Introduction By a letter dated August 2,1996, as supplemented July 1,1997 Duquesne Light Company (DLC) submitted for staff review a copy of the " Beaver Valley Unit 1, 1996 Alternate Repair Criteria 90 Day Report." The report was submitted in accordance with Generic letter (GL) 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," as a result of implementing the voltage-based alternate repair criteria in the Beaver Valley Power Station, Unit No.1 (BVPS-1) technical specifications (TSs).

GL 95-05 allows steam generator tubes having outside diameter stress corrosion cracking (00 SCC) that is predominately axially oriented and confined within the tube support plates to remain in service on the basis of, in part, bobbin coil voltage response. GL 95-05 specifies that inspection results and associated tube integrity analyses should be submitted within 90 days of each plant restart followin? a steam generator tube inspection. The report should include, at a minimum, calculations of the voltage distribution, postulated tube leakage, and tube burst probability under normal and accident conditions.

2.0 General Plant Descriotion BVPS-1 has three Westinghouse Model 51 steam generators. The tubes are seven-eighths of an inch in diameter and were fabricated with mill annealed alloy 600 material. The steam generators have drilled hole tube support plates.

On April 1,1996, the NRC staff approved the licensee's request to implement the 2-volt alternate repair criteria for cycle 12 in the BVPS-1 TSs.

Subsequently.-the licensee performed the end-of-cycle 11 (EOC-11) tube inspection during the spring 1996 outage. During the EOC-11 inspection, plugs were removed from previously repaired tubes, the tubes were reinspected, and those with indications satisfying the 2-volt alternate criteria were returned to service in the beginning of cycle 12. Cycle 12 will be completed in September 1997.

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3.0 Insoection Scone and Results The licensee used the bobbin coil probe to examine all tube support plate intersections in the tube bundles of all three steam generators. The rotating pancake coil (RPC) was used to inspect bobbin indications with amplitudes i greater than 2 volts at the tube support plate intersections. The licensee detected a total of 1936 indications at tube support plate intersections i during the E0C-11 inspection. Of those, 83 indications were removed from service, of which only two indications did not meet the 2-volt repair criteria

! (the remaining indications were plugged for causes other than 00 SCC at tube I . The remaining 1853 indications were returned to service for

! support cycle 12. plates)ddition, In a 508 indications from deplugged tubes were recovered i and returned to service. This resulted in a total of 2361 indications returned to service for cycle 12.

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! 4.0 Voltana Distribution Calculations t

j To project voltage distributions of indications for the E00-12, the licensee used the methodology in the Westinghouse report, WCAP-14277 (Reference 1) j . The NRC staff j which performed was approved calculation a confimatory in an NRCbased safety on GL evaluation (Reference 2)dologies 95-05 metho and fo 1 that the projected end-of-cycle voltage distribution agreed with that calculated by the licensee. Based on this comparison, the NRC staff concludes that the licensee's methodology for calculating voltage distribution appears

< to be consistent with the methodology outlined in GL 95-05.

j 5.0 Conditional Probability of Burst Calculations l The licensee projected that the conditional probability of burst during a main

steam line break (MSLB) at the EOC-12 for steam generators A, B, and C would 4 be 1.23E-4, 1.76E-4, and 1.02E-4, respectively. The NRC staff independently

! calculated conditional probability of burst for three steam generators for the EOC-12. The NRC staff's results were found to be in close agreement with the j licensee's results. In addition, the projected conditional probability of 4

i burst for the EOC-12 is well within the limit of 1.0E-2 as defined in GL 95-05.

6.0 Proiected MSLB Leakaae Calculations For the E0C-12, the licensee projected MSLB leak rates of 4.28, 4.5, and 2.76 gallons per minute (gpm) for steam generators A, B, and C, respectively. The projected MSLB leak rates are within the allowable limit of 4.5 gps, which was derived in accordance with the guidelines of 10 CFR Part 100 and General-Design Criterion 19. The NRC staff also independently calculated MSLB leak rates for the E0C-12. The NRC staff's results were found to be in close agreement with the licensee's projected leak rates.

During a conference call that took place on September 2,1997, the NRC staff was informed that the licensee had properly accounted for temperature density effects in comparing calculated leakage and site allowable leakage.

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3 i 7.0 Ca=aarison of Proincted EDC-11 Conditions to As-Found EOC-11 Conditions As a part of the 90-day report review, the NRC staff compared the projected EOC-11 results to that of the as-found E0C-Il results obtained during the inspection. The purpose of the comparison was to verify the ability of the predictive methodoloflies to provide a conservative projection of the number and distribution of 'ndications at the next [0C such that the estimated conditional probability of burst and total leak rate under postulated MSLB conditions at the next E0C are conservative.

The largest predicted voltages were about the same as the largest measured voltages. The licensee over-predicted the number of indications greater than 0.7 volts and under-predicted the number of indications less than 0.7 volts.

The inadequate prediction of the number of indications less than 0.7 volts did not significantly affect the projected conditional probability of burst or ,

i leak rate because the number of indications in the higher voltage bin were conservatively projected. The licensee under-predicted the projected  ;

1 conditional probability of burst for steam generators A and 8, and over-predicted for steam generator C. Despite the under-predictions, the licensee's projected burst probabilities (e.g., 5.0 to 6.5E-5) remain much below the G. 95-05 limit of 1.0E-2. The difference between the projected and actual burst probabilities is not meaningful given the very low conditional

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probability of burst calculated. With respect to leakage calculations, the licensee over-predicted the EOC-11 leak rates for all taree steam generators.

The leak rates were all less than the allowable of 4.5 gpm 8.0 Tube Pull Result DLC did not pull any tubes during the EOC-Il outage. According to the schedule prescribed in GL g5-05, the licensee was not required to remove any tubes for destructive examination during the E00-11 outage, g.0 Pronosed Alternative Methods In addition to the GL 95-05 recommended calculations, the licensee performed additional calculations based on three proposed alternative methods that deviate from current NRC-approved methods. They are: (1) probability of detection based on voltage and prior cycle detection instead of a constant value of 0.6; (2) the use of best estimates for leak rates instead of the log-normal distribution for leak rate; and (3 data from French nuclear plants in the cur) rent seven-eighths of an inchcalculations with th diameter tube database removed. The NRC's staff is reviewing these three-proposed. alternative methods, which are discussed in a report (Reference 5) submitted by the Nuclear Energy Institute, as a part of generic industry initiatives. The resolution of the licensee's proposed alternative methods is dependent on the resolution of the generic review by the NRC staff.

e 4-10.0 Suuntry The comparison of the projected E00-11 results to that of the as-found EOC-11 results indicate that predictive methodologies are yielding good results. In addition, the calculated EOC-11 conditional probability of burst and projected MSLB limits, leak rate were less than GL 95-05 criteria and site allowable leak rate respectively.

Lastly, the projected EOC-12 conditional probability of <

burst and projected MSLB leak rate were also less than the GL 95-05 criteria and site allowable leak rate limits, respectively.

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. l 11.0 References

1. WCAP-14277, 'SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections,' Westinghouse Nuclear Services Division, January 1995.

2 ' Safety Evaluation by the Office of Nuclear Reactor Regulation Related to l Amendment No. 198 to Facility Operating License DPR-66, Duquesne Light l

Company, Ohio Edison Company and Pennsylvania Power Company, Beaver Valley  !

Power Station, Unit No.1 Docket No. 50-334 ' U.S. Nuclear Regulatory Commission April 1,1996.

3. WCAP-14123, ' Beaver Valley Unit 1 Steam Generator Tubs Pluoring Criteria for Indications at Tube Support Plates," Westinghouse Elect.ic Corporation, Proprietary Class 2, July 1994.

4. NP-7480-L, " Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits - 1996 Database Update,' Addendum 1, Electric Power Research Institute, June 1996.

5. NP-7480-L, " Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits - 1996 Database Update,' Addendum 1 Electric Power Research Institute, November 1996 e = =-%