ML20217Q630
| ML20217Q630 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 04/08/1998 |
| From: | Alexion T NRC (Affiliation Not Assigned) |
| To: | Cottle W HOUSTON LIGHTING & POWER CO. |
| References | |
| GL-95-05, GL-95-5, TAC-MA0966, TAC-MA966, NUDOCS 9804130052 | |
| Download: ML20217Q630 (6) | |
Text
Mr. Willi m T. Cottla April 8, 1998 President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 l
Wadsworth,TX 77483 J
SUBJECT:
STEAM GENERATOR TUBE VOLTAGE-BASED REPAIR CRITERIA 90-DAY j
REPORT, SOUTH TEXAS PROJECT, UNIT 1 (TAC NO. MA0966)
Dear Mr. Cottle:
By letter dated December 18,1997, the South Texas Project Nuclear Operating Company (STPNOC) submitted a steam generator voltage-based repair criteria 90-day report, for South Texas Project, Unit 1 (STP-1). The report summarized the results of the STPNOC's condition monitoring and operational assessments for the unit's steam generators in accordance with NRC Generic Letter 95-05.
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The STP-1 Technical Specifications include a reporting threshold of 1x10 for the conditional probability of tube burst as part of the implementation of the voltage-based tube repair criteria.
1 STPNOC estimated a conditional burst probability below this threshold using NRC staff-approved methodology. The projected primary-to-secondary leak rates during a postulated steam line break event for each of the STP-1 steam generators were significantly below the maximum site allowable accident leak rate of 5.0 gallons per minute and were determined by STPNOC using NRC staff-approved methodology.
l Based on its review, the NRC staff concludes that STPNOC implemented the voltage-based repair criteria in accordance with its licensing basis. The staff's review of the licensee's 90-day outage reporiis enclosed.
j Sincerely, ORIGINAL SIGNED BY:
Thomas W. Alexion, Project Manager Project Directorate IV-1.
Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation Docket No. 50-498 1
Enclosure:
90-Day Report Review cc w/ encl: See next page DISTRIBUTION:
Docket File PUBLIC PD4-1 r/f EAdensam (EGA1)
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UNITED #TATES j
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i April 8, 1998 l
Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth, TX 77483
SUBJECT:
STEAM GENERATOR TUBE VOLTAGE-BASED REPA!R CRITERIA 90-DAY REPORT, SOUTH TEXAS PROJECT, UNIT 1 (TAC NO. MA0966)
I
Dear Mr. Cottle:
By letter dated December 18,1997, the South Texas Project Nuclear Operating Company (STPNOC) submitted a steam generator voltage-based repair criteria 90-day report, for South Texas Project, Unit 1 (STP-1). The report summarized the results of the STPNOC's condition monitoring and operational assessments for the unit's steam generators in accordance with NRC Generic Letter 95-05.
The STP-1 Technical Specifications include a reporting threshold of 1x10-2 for the conditional probability of tube burst as part of the implementation of the voltage-based tube repair criteria.
STPNOC estimated a conditional burst probability below this threshold using NRC staff-approved methodology. The projected primary-to-secondary leak rates during a postulated steam line break event for each of the STP-1 steam generators were significantly below the maximum site allowable accident leak rate of 5.0 gallons per minute and were determined by STPNOC using j
l NRC staff-approved methodology.
Based on its review, the NRC staff concludes that STPNOC implemented the voltage-based repair criteria in accordance with its licensing basis. The staffs review of the licensee's 90-day outage report is enclosed.
Sincerely, WLW
)
=
Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects lil/lV Office of Nuclear Reactor Regulation j
Docket No. 50-498
Enclosure:
90-Day Report Review cc w/ encl: See next page
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[
Mr. William T. Cottle STP Nuclear Operating Company South Texas, Units 1 & 2 cc:
l Mr. David P. Loveless Jack R. Newman, Esq.
l Senior Resident inspector Morgan, Lewis & Bocklus U.S. Nuclear Regulatory Commission 1800 M Street, N.W.
P. O. Box 910 -
Washington, DC 20036-5869 Bay City, TX 77414 Mr. Lawrence E. Martin A. Ramirez/C. M. Canady Vice President, Nuc. Assurance & Licensing City of Austin STP Nuclear Operating Company.
Electric Utility Department P. O. Box 289 721 Barton Springs Road Wadsworth,TX 77483
. Austin, TX 78704.
Office of the Govemor Mr. M. T. Hardt ATTN: John Howard, Director l
Mr. W. C. Gunst Environmental and Natural l
City Public Service Board Resources Policy l
P. O. Box 1771 P. O. Box 12428 San Antonio, TX 78296 Austin, TX 78711 Mr. G. E. Vaughn/C. A. Johnson Jon C. Wood Central Power and Light Company Matthews & Branscomb P. O. Box 289 One Alamo Center.
t Mail Code: N5012 106 S. St. Mary's Street, Suite 700 Wadswodh,TX 74483 San Antonio, TX 78205-3692 INPO.
Arthur C. Tate, Director Records Center Division of Compliance & Inspection 700 Galleria Parkway Bureau of Radiation Control Atlanta, GA 30339-3064
' Texas Department of Health 1100 West 49th Street i
Regional Administrator, Region IV '
Austin, TX 78756 U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Jim Calloway Artington, TX 76011 Public Utility Commission of Texas l
Electric Industry Analysis D. G. Tees /R. L Balcom P. O. Box 13326 l
Houston Lighting & Power Co.
Austin, TX 78711-3326 P. O. Box 1700 l:
- Houston, TX 77251
~ Judge, Matagorda County l
Matagorda County Courthouse L
1700 Seventh Street L
Bay City, TX 77414
4 STEAM GENERATOR 90-DAY REPORT REVIEW i
By letter dated December 18,1997, the South Texas Project Nuclear Operating Company (the licensee) submitted the "1RE07 Steam Generator Tube Voltage-Based Repair Criteria 90-Day Report," for South Texas Project, Unit 1 (STP-1). This report summarizes the licensee's condition monitoring and operational assessments for Cycle 7 and 8 operation, respectively. The j
staff has reviewed the licensee's 90-osy report per the critoria included in NRC Generic Letter (GL) 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking." The following documents the staffs j
conclusions from the review of this report.
1.0 GENERAL PLANT DESCRIPTION 4
JTP-1 has four Westinghouse Model E2 steam generators with 3/4-inch diameter tubes. The licensee implemented the 1.0 volt steam generator tube attemate repair criteria for the first time in the outage prior to 1RE07 (1RE06, May 1996). STP-1 has operated one full operating cycle since the initialimplementation of the voltage-based repair criteria.
2.0 -
STEAM GENERATOR TUBE EDDY CURRENT INSPECTION SCOPE AND RESULTS The licensee inspected 100% of the STP-1 steam generator tubes over their full length using a 0.610-inch diameter bobbin coil. The bobbin coil probe examinations identified 1151 indications at tube support plate (TSP) intersections. Rotating pancake coil (RPC) examinations of bobbin indications greater than one volt are required by the voltage-based repair criteria. A total of 15 TSP indications were inspected with an RPC probe. Nine of the 15 indications measured greater than one volt with a bobbin coil probe. Three TSP indications of outside diameter stress corrosion cracking (ODSCC) were confirmed during the RPC inspections and removed from service.
Based on the number of tubes reiumed to service with bobbin coilindications, steam generator C is considered to be the limiting steam generator for Cycle 8 operation. As discussed in Section 3.3 of this report, this conclusion is supported by the calculated estimate of accident tube leakage.
3.0 EVALUATION OF PROBABILISTIC METHODOLOGIES FOR ESTIMATING CONDITIONAL PROBABILITY OF BURST AND TOTAL LEAK RATE UNDER l
POSTULATED STEAM LINE BREAK CONDITIONS l
Acceptable tube integrity at the conclusion of Cycle 8 operation is demonstrated, in part, by a l
calculated conditional orobability of tube burst for the limiting steam generator less than the
' reporting threshold indicated in GL 95-05 and an estimated accident-induced steam generator tube leak rate from ODSCC at tube support plate intersecti'ons below plant-specific limits. Three distinct probabilistic calculations are necessary to determine these results. The following summarizes the staffs evaluation of the results reported for these calculations.
. 3.1 Projected End-of-Cycle Voltage Distribution
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The licensee's 90-day report compared the as-found distribution of voltages determined from the EOC-7 refueling outage to those estimated from the predictions made following the prior outage ENCLOSURE
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inspections (EOC-6). Projechons of the distribution of voltages made following the EOC-6 inspections estimated that the four STP-1 steam generators would contain approximately 1700
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tube indications in the EOC-7 outage. Steam generator "C" was considered to be the limiting d
- steam generator with respect to the overall number and voltage of indications projected. A comparison between the as-found distribution and the projected distribution of voltages indicates that the predictive methodology previously used to estimate the voltages of indications in
. subsequent outages yields conservative results. The maximum predicted indication voltage was estimated to be approximately 2.1 volts. However, the maximum voltage identified during the inspection was 1.8 volts. The maximum indication voltages identified within the other three -
steam generators during the EOC-7 inspection were also less than or equal to the previous outage projections. In addition, the number of indications identified in each steam generator was less than the previous estimates.
In order to obtain the most conservative results with respect to the growth rate distribution used in the Monte Carlo simulation, the licensee utilized both steam generator specific and hybrid growth rate distributions obtained from operation in Cycle 6. Steam generator specific growth rates were applied to steam generators "B" and "D". Composite growth distributions developed from all tubes in service _ during Cycle 6 were applied to steam generators "A" and "C". Growth
- rate distributions included in the 90-day report were reported on an effective full-power year-(EFPY) basis. For the determination of EOC-8 voltage distributions, the growth rates were scaled to the estimated length of operation over the next cycle,1.39 EFPY.
Using the inspection findings in the EOC-7 inspection, the licensee calculated the projected EOC-8 voltage distribution for TSP indications. Based on the overall number of indications anticipated at the EOC-8, the "C" steam generator is the limiting steam generator for the next cycle of operation. However, the distribution of higher voltage indications for the "A" and *C" steam generators is similar. Therefore, either of these two steam generators could be limiting from a burst or leakage perspective at EOC-8. The staff independently verified the licensee's calculations by completing Monte Cado simulations to estimate the EOC-8 voltage distributions.
The results of these calculations confirm that the predictive methodology used by the licensee to L estimate the EOC voltage distributions is consistent with the guidance provided in GL g5-05.
3.2 Conditional Probability of Tube Burst Due to the relatively loyv number of higher voltage indications both projected and identified during inspections at STP-1, the calculated conditional probability of tube burst is typically below the reporting threshold of 1x10-2 specified in GL g5-05. Following the EOC-6 refueling outage, the 4
licensee estimated an.EOC-7 burst probability for the limiting steam generator of 7.5x10. Using the actual inspection results as the input into the calculation, the licensee determined the as-found conditional burst probability following the most recent inspections. The resulting burst probability is 3.1x10-5. Therefore, the calculational methodology used to predict the EOC-7 i
conditional burst probabilities yielded conservative results for this steam generator over the previous cycle of operation.
Using the growth rate distributions determined for each steam generator, the licensee reported the projected EOC-8 conditional tube burst probabilities. The ca:culated probability of tube burst for steam generator "A" (9.4x10*) slightly exceeded that determined for the "C" steam generator
' (8.3x10~S). All projected values are well below the GL repor1ing threshold, and therefore, the' estimated tube burst probability due to ODSCC at tube support plates is well within acceptable limits for Cycle 8 operation.-
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3.3 Steam Line Break Leak Rate Projection l
The staff evaluated the steam line break leak rate reported by the licensee similar to the l
assessment of the conditumal tube burst probabilities.' For the limiting steam generator ('C"), the l-EOC-7 projected leak rate was previously estimated to be 7.5x10 gallons per minute (gpm) '
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versus an as-found estimate based on the inspechon results obtained in the EOC-7 outage of 4
9.7x10 gpm. Due to the high number of indications projected for steam generator "C", it is i
considered the limiting steam generator from a leakage perspective for the Cycle 8 operation.
The estimated steam line break leak rate for this steam generator for the EOC-8 is 5.3x10 gpm.
4 This value is several orders of magnitude below the STP-1 steam line break leak rate limit of 5.0 l
gprn. Therefore, the projected tube leakage integrity for ODSCC indications is well within the allowable limit established for STP-1.
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. DATABASE FOR TUBE INTEGRITY CALCULATIONS I
in order to calculate the conditional tube' burst' probabilities and postulated steam line break l
primary-to-secondary leak rate, the methodology approved for GL 95-05 requires the use of burst l
and leak rate data obtained from model bo!Ier tubes and tubes removed from actual steam.
generators. The industry has developed correlations relating bobbin coil voltage to the measured j
leak rate, probability of burst, and burst pressure through testing of these tubes. The licensee's 90-day report included the mathematical expressions for each of these correlations. The staff l~
has compared the parameters of these correlations with those developed using data which i
. should currently be included in the database for 3/4-inch tubing. The parameters were the same j
l for each of the three correlations. Therefore, the staff concludes that the licensee employed the l
appropriate data in the tube integrity calculations for STP-1.
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l 5.0 TUBE PULL RESULTS l
The licensee for STP-1 removed steam generator tubes with ODSCC indications during the plant l
outage in the spring of 1995. Per the guidance provided in GL 95-05, tube pulls are not required i
L until the outage that begins after January 1998.. Therefore, at the next refueling outage, the licensee would be expected to remove additional tube samples to confirm the nature of the degradation occurring in the STP-1 steam generators and augment the industry ODSCC j
database through burst and leak testing of the samples.
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SUMMARY
Based on its review, the NRC staff concludes that STPNOC implemented the voltage-based l
t repair criteria in accordance with its licensing basis.
Principal Contributor. P. Rush Date: April 8. 1998 l
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