ML17229A368
| ML17229A368 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 06/03/1997 |
| From: | Stall J FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML17229A369 | List: |
| References | |
| RTR-NUREG-0800, RTR-NUREG-800 GL-95-05, GL-95-5, L-97-141, NUDOCS 9706090306 | |
| Download: ML17229A368 (8) | |
Text
CATEGORY 1 REGULATO INFORMATIPN DISTR'IBUTION TEM (RIDS)
ACCESSION NBR:9706090306 DOC.DATE: 97/06/03 NOTARIZED: NO DOCKET ¹ FACIAL:50-.335 St. Lucie Plant, Unit 1, Florida Power & Light Co. 05000335 AUTH.NAM" AUTHOR AFFILIATION S;.ALL,J.A. Florida Power.& Light Co.
RECIP.NAME RECIPIENT AFFILIATION oe/
Document Control Branch (Document Contro Desk)
SUBJECT:
Forwards Rev 1 to SAIC-97/1008, "Analysis of Radiological Consequences of Main Steam Line Break Outside Containment for St Lucie Unit 1 Nuclear Power Plant Using NUREG-0800 Std Review Plan Section 15.1.5,App A."
DISTRIBUTION CODE: AOOID TITLE: OR COPIES RECEIVED:LTR )
Submittal: General Distribution ENCI j SIZE:Q +DQ NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-3 LA 1 1 PD2-3 PD 1 1
, WIENSiL. 1 1 INTERNAL: ACRS 1 1 FILE CENTER 1 1 NRR/DE/ECGB/A 1 1 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1, NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 '1 NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP"US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXTi 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL .NUMBER OF COPIES REQUIRED: LTTR 14 ENCL 13,
Florida Power & Light Company, 6501 South Ocean Orive, Jensen Beach, FL 34957 June 3, 1997 L-97-141 10 CFR 50.4 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 RE: St. Lucie Unit 1 Docket No. 50-335 Dose Assessment Correction The purpose of this letter is to provide a revision to the dose assessment submitted by Florida Power "8c Light Company (FPL) on April 4, 1997, and to provide additional information on the Fall 1996 steam generator inspection results. Attachment 1 provides the requested clarifications and additional information'on the FPL steam generator inspection results for St. Lucie Unit 1.
Attachment 2 provides the revised St. Lucie Unit 1 dose assessment using Standard Review Plan Section 15.1.5 methodology.
The revised dose assessment, prepared by Scientific Applications International Corporation (SAIC) for FPL, is SAIC Report, SAIC-97/1008 Revision 1, Analysis of the Radiological Consequences of a Main Steam Line Break Outside Containment for the St. Lucie Unit 1 Nuclear Power Plant Using NUREG-0800 Standard Review Plan 15.1.5 Appendix A, dated April 29, 1997. Revision 1 of the SAIC report incorporates a radionuclide release scenario which provides more realistic postulated doses when compared to doses estimated in Revision 0,'.which was provided as Attachment 2 to our April 4, 1997 submittal (L-97;90). The revised report provides a SRP dose calculation for an assumed pre-accident one gpm primary-to-secondary leak and a variable, but constant, post-accident primary-to-secondary leak rate for a postulated main steam line break (MSLB) outside containment. The three cases which were analyzed are: (1) a preexisting iodine spike in the primary coolant system, (2) an accident induced iodine spike, and (3) an accident induced 1.61 percent (%) fuel failure. The bounding results of this analysis are tabulated below and show the maximum allowable post-accident primary-to-secondary leak rate required to reach the regulatory dose limits set forth in 10 CFR 100 and 10 CFR 50, Appendix A, General Design Criteria 19, as indicated in SRP Section 15.1.5, Appendix A. Since the thyroid dose is more limiting than the whole body dose, only the limiting results are presented.
These results show that, for the bounding case of a postulated MSLB with induced steam generator tube leakage following 15 months of operation, the calculated upper bound leakage of less than 4 gpm from all forms of degradation, is substantially below the 6.8 gpm leak rate which is considered acceptable under the pending steam generator rule making criteria, in that the results do not exceed the 10 CFR 100 and 10 CFR 50 Appendix A general design criteria (GDC) 19 accident dose limits.
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St. Lucie Unit 1 Docket No. 50-335 L-97-141 Page 2 Regulatory Thyroid Dose Limiting Post-Accident Primary-to-Secondary Leak Rate Dose Location Regulatory Concurrent 1.61. % Failed Thyroid Dose Iodine Spike Iodine Spike Fuel Case 300'reexisting Limit (REM) Case (gpm) Case (gpm) (gpm)
Control Room 30 67 196 6.8 Exclusion Area Boundary 30 or 430 157 33 300'0 (Site Boundary)
Low Population Zone or 3,000 988 LPZ For the concurrent iodine spike case, the regulatory limit is 30 REM; for the other cases the regulatory dose limit is 300 REM During a July 3, 1996, meeting with the NRC, Florida Power and Light Company (FPL) committed to provide the results of the St. Lucie Unit 1 steam generator run time analysis to the NRC within 90 days of the startup from the 1996 refueling outage (SL1-14). FPL letter L-96-273 dated October 24, 1996, provided FPL's plans to operate the Unit 1 steam generators for fifteen (15) months (i.e., through October 23, 1997). FPL letters L-97-47 dated February 21, 1997, and L-97-90 dated April 4, 1997, supplemented this information in response to an NRC RAI dated January 23, 1997. The run time analysis was a physically based analysis that used the guidance contained in Draft Regulatory Guide (RG) 1. 121, Generic Letter 95-05, and the Draft Regulatory Guide, Steam Generator Tube Integrity.
This letter does not contain any new regulatory commitments. FPL is prepared to meet with the NRC to discuss the results of this and any other previously submitted analyses. Please contact us ifyou have any additional questions.
Very trul yours
. A. Stall Vice President St. Lucie Plant JAS/GRM Attachments cc: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant
St. Lucie Unit 1 Docket No. 50-335 L-97-141 Attachment 1 Page 1 Cycle 14 Steam Generator Runtime Analysis I3d Jl p NRC Question 1:
Clarify the following statement on page 60 of Enclosure 2 of your October 24, 1996 submittal, "St. Lucie Unit 1; Docket No. 50-335; Steam Generator Run Time Analysis for Cycle 14" "The number of simulated leaking defects in the entire steam generator is obtained by first sampling from a binomial distribution using the average level frequency of leak computations and the number of equivalent levels in the steam generator?"
FPL Response to Question 1:
The number of simulated leaking defects for the entire steam generator is obtained in a two-step process. The first step is the computation of the number of leaking defects at an average level. As in the structural integrity portion of the analysis (Section 5.1), the 'average'evel corresponds physically to roughly one tube support level. The number of leaking defects for this average support is obtained by sampling from a Binomial distribution using the probability of any leaking defect in the average support. The result of this process is a trial outcome of N leaking defects (N= 0,1,2,3,.....,k) for the average support.
The second step in the computation is the summation of leaking defects over all levels in the steam generator. As in Section 5.1 of the report, the number of such levels is 15 for the most affected steam generator. The summation process is ordinary addition unlike that for probability of burst which is a Boolean summation.
NRC Question 2:
In Table 4-1 of Enclosure 4 to your October 24, 1996 submittal, the maximum bobbin depth for tube R15L55 at 1H is recorded as "94/64." Please clarify what is meant by this entry.
FPL Response to Question 2:
Table 4-1 of Enclosure 4 to our October 24, 1996 submittal provides results of In-situ Pressure Tests conducted for defect indications in St. Lucie Unit 1 steam generator tubing. These results were discussed with the Staff at the completion of the testing. Prior to in-situ pressure testing, all test candidate indications were reviewed by lead analysis personnel. The maximum bobbin
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St. Lucie Unit 1 Docket No. 50-335 L-97-141 Attachment 1 Page 2 depth for R15L55 at 1H was subsequently listed as "94/64" due to the complex nature of the indication. Theoverallindicationismeasuredas64% through wall. A smallercomponentof the indication, however, when measured separately, measures 94% through wall. In-situ testing for this indication demonstrated that, while leakage did occur, adequate structural margins existed.
The specific entry (94/64) reflects this result.
NRC Question 3:
In Hgures 13 and 14 of Appendix A to Enclosure 1, the number of indications with depths between 0-19%, 20-39%, and ) 40% were provided. Please provide a breakdown of the number of indications between 40-50%, 50-60%, ... and 90-100% for the indications greater than 40% in 1996. Also, please state whether the depths are max depth or average depth.
Ifnot max depth please provide max depth readings.
FPL Response to Question 3:
A breakdown of the number of indications between 40-50%, 50-60%, ... and 90-100% for indications greater than 40% in 1996 is provided in Table 1 for steam generator A and Table 2 for steam generator B.
Data analysis personnel were instructed to provide maximum eddy current depth estimates for all indications. At St. Lucie Unit 1, bobbin coil indications at eggcrates, that are 2.5 volts or greater in amplitude, may exhibit a complex signal indicative of an indication with variable depth. As discussed in our response to Question 2, such an indication at R15L55 at 1H in steam generator B was noted during reviews prior to in-situ pressure testing. This indication is included in Table 2 as 64%, although a smaller component of the indication was later measured as 94% through wall. A sample review of bobbin data verified that eggcrage indications less than 2.5 volts in amplitude do not exhibit such complex signals. In addition, all remaining eggcrate indications measuring 2.0 volts or greater were reviewed and no additional exceptions were noted.
St. Lucie Unit 1 Docket No. 50-335 L-97-141 Attachment 1 Page 3 Table 1 St. Lucle Unit 1 - SGA1996, >40% Depth Breakdown Elevation NQ1 40A9% 50-59% 60-69% 70-79% 80-89% 90%+ TOTALS 01 H 84 41 27 15 3 2 172 02H 46 47 15 2 2 1 113 03H 36 20 8 67 04H 20 13 2 2 1 38 05H 21 6 4 1 32 06H 8 2 2 12 07H 7 7 08H 1 09H 0 10H 0 01 C 9 02C 8 03C 8 04C 9 05C 5 06C 2 07C 7 08C 1 09C 1 10C 0 VS 1 2 VS 2 6 VS 3 0 TSH 37 28 73 TEH 0 TS C 0 TEC 0 DC 8 0 DHB 0 DHT 0 DCT 0 TOTALS 302 164 73 24 573 NQI - Non Quantifiable Indication 01H, 02H .... - 1st support hot side 01C, 02C .... - 1st, 2nd support cold side VS1,'S2.... - 1st, 2nd Vertical Strap from hot side DHB, DCB - Diagonal Strap, Hot or Cold Side, Bottom Edge DHT, DCT- Diagonal Strap, Hot or Cold Side, Top Edge TSH, TSC - Tubesheet Hot or Cold, Secondary face TEH, TEC - Tube End Hot or Cold Side
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St. Lucie Unit 1 Docket No. 50-335 L-97-141 Attachment 1 Page 4 Table 2 St. Lucie Unit 1 - SGB 1996, >40% Depth Breakdown Elevation NQI 40-49% 50-59% 60-69% 70-?9% 80-89% 90%+ TOTALS 01 H 173 37 31 16 17 5 279 02H 80 16 7 108 03H 40 10 7 2 3 62 04H 22 5 4 31 05H 20 21 06H 6 2 2 10 07H 4 1 2 7 08H 6 1 1 8 09H 0 10H 0 01 C 10 11 02C 2 7 03C 5 6 04C 5 8 05C 3 6 06C 6 8 07C 5 7 08C 4 5 09C 0 10C 1 VS 1 9 9 VS 2 17 18 VS 3 1 1 TS H 26 10 42 TEH 0 TS C 4 TEC 0 DCB 0 DHB 0 DHT 0 DCT 0 TOTALS 447 97 60 27 21 659 NQI - Non Quantifiable Indication 01H, 02H .... - 1st support hot side 01C, 02C .... - 1st, 2nd support cold side VS1, VS2.... 1st, 2nd Vertical Strap from hot side DHB, DCB - Diagonal Strap, Hot or Cold Side, Bottom Edge DHT, DCT - Diagonal Strap, Hot or Cold Side, Top Edge TSH, TSC - Tubesheet Hot or Cold, Secondary face TEH, TEC - Tube End Hot or Cold Side