ML20217N513
| ML20217N513 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/30/1998 |
| From: | Mccoy C SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCV-1190, NUDOCS 9804090121 | |
| Download: ML20217N513 (10) | |
Text
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C.K.McCoy Southern Nuclear Vice President Operedne Company. Inc.
Vogtle Project 40 inverness Center Parkway P.0, Box 1295 Birmingham, Alabama 35201 Tel 205.992.7122
.s Fax 205.992.0403
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SOUTHERN h COMPANY Harch 30, 1998 Energy to Serve hurWorld" LCV-1190 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
VOGTLE ELECTRIC GENERATING PLANT 10CFR50.46 ECCS EVALUATION MODELS 1997 ANNUAL REPORT Attached is Southern Nuclear Operating Company, Inc.'s (Southern Nuclear) 10CFR50.46 Emergency Core Cooling System (ECCS) Evaluation Models 1997 Annual Report based on WCAP-13451 and in compliance with the reporting requirements of 10CFR50.46(a)(3)(ii). It is based on information provided by Westinghouse of errors and changes assessed against the Vogtle Electric Generating Plant (VEGP) ECCS Evaluation Models since the 1996 Annual Report.
The attached report summarizes the effects of changes and errors in the ECCS Evaluation Models on peak clad temperature (PCT). Also, the report provides a summary of the plant change safety evaluations performed under the provisions of 10CFR50.59 that also affect PCT. The report results will be incorporated into the next Final Safety Analysis Report (FSAR) update.
As shown in the attached 1997 Annual Report, compliance with the PCT requirements of 10CFR50.46 continues to be maintained when the effects of plant design changes are
/
combined with the effects of the ECCS Evaluation Models assessments applicable to
[
VEGP Units I and 2.
Ifyou have any questions regarding this report, please contact this office.
gp /
Sincerely, 9804090121 980330 PDR ADOCK 05000424 S
PDR C. K. McCo CKM/ RJF/gmb Attachment
4 U. S. Nuclear Regulatory Commission Page 2 cc:
Southern Nuclear Ooerating Comoany Mr. J. B. Beasley Mr. M. Sheibani -
NORMS U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. D. H. Jaffe, Senior Project Manager, NRR Mr. J. Zeiler, Senior Resident inspector, Vogtle i
LCV-Il90
ATTACHMENT VOGTLE ELECTRIC GENERATING PLANT 10 CFR 50.46 ECCS EVALUATION MODELS 1997 ANNUAL REPORT BACKGROUND Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models on an annual basis when the errors and changes are not significant, and within 30 days ofdiscovery when the errors and changes are significant A significant error or change, as defined by 10 CFR 50.46, is one which results in a calculated fuel peak cladding temperature (PCT) different by more than 50*F from the temperature calculated for the limiting transient using the last acceptable model, of a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50*F, The following presents a summary of the effects of errors and changes to the Westinghouse ECCS Evaluation Models on the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 loss of coolant accident (LOCA) analyses since the 1996 annual report (Reference 1). This annual -
report has been prepared in accordance with the methodology presented in WCAP-13451 (Reference 2). The LBLOCA and SBLOCA analyses, Evaluation Model assessments, and safety evaluation results reported herein will be included in a future VEGP Final Safety Analysis Report (FSAR) update.
LARGE-BREAK LOCA ECCS Evaluation Model Since the previous report (Reference 1), no new assessments against the VEGP LBLOCA analysis have been identified. The LBLOCA analysis results are based on the Westinghouse IIASH large-break ECCS Evaluation Model (Reference 3), as approved by the NRC for VEGP-specific application (References 4 and 5) and the latest acceptable LOCBART model.
The limiting size break analysis continues to assume the following information important to the LBLOCA analyses:
o 17x17 VANTAGE-5 Fuel Assembly o
Core Power = 1.02
- 3565 MWT o
Vessel Average Temperature = 571.9"F o
Steam Generator Plugging Level = 10%
o Fo = 2.50 o
FAH = 1.65 l
ATTACHMENT Page 2 For VEGP Units 1 and 2, the limiting size break continues to be the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of Co== 0.6. The LBLOCA LOCBART analysis-of-record calculated PCT value is 1915T as discussed below.
In the 1996 report (Reference 1), Southern Nuclear reported a LOCBART Reanalysis Result
- PCT value of 191IT, which reflected a -4T change from the previously reported value of 19157 (Reference 9). The change of-4*F was for the LOCBART clad creep and burst error.
Southern Nuclear has decided to track the LOCBART clad creep and burst error as a separate item under Analysis-of-Record rather than combining it into the LOCBART Analysis Result.
Therefore, the LOCBART Analysis Result is reported as 19157, and the Analysis-of-Record
- category includes a separate -4*F for the LOCBART clad creep and burst error.
' ' LThe containment purge, T,3 uncertainty, and transition core penalty items continue to be listed separately per the format of WCAP-13451. The items are listed separately because these items are not explicitly modeled. The PCT assessment values on these items remain 10,11, and 507, respectively. The VEGP Unit 2 core contained all VANTAGE-5 fuel during 1997.'
Therefore, the Transition Cycle Penalty did not apply to VEGP Unit 2.
VEGP began using ZIRLO clad fuel rods in the Unit 2 Cycle 6 and Unit 1 Cycle 8 cores. The use of 2IRLO clad fuel rods results in a penalty of 5*F PCT as calculated by the latest -
acceptable LOCBART model.
VEGP is using ZIRLO clad IFBA fuel rods with a backfill pressure of 100 psig in the Unit 2 Cycle 6 and Unit 1 Cycle 8 cores. The ZlRLO clad IFBA rods result in a penalty of 21T PCT l
as calculated by the latest acceptable LOCBART model.
Because there is a mix of Zircaloy and ZIRLO clad fuel rods and IFBA and non-IFBA rods in VEGP core designs, VEGP will continue to show an analysis-of-record LOCBART calculated PCT value based on non-IFBA, Zircaloy fuel rods (1915T) and will apply PCT penalties when the VEGP cores contain the ZIRLO clad fuel rod.
Prior BASH Larne-Break ECCS Evaluation Model Assessmenu i
In the following table and as reported in Reference 1, four prior model assessments have been combined into a single assessment of-6*F. These assessments am: (1) Steam G aerator Flow Area Application, (2) Structural Metal Heat Modeling, (3) LUCIFER Error Correction, and
- (4) Translation of Fluid Conditions from SATAN to LOCTA.
ATTACHMENT Page 3.
i 1997 BASH Large-Break ECCS Evaluation Model Assessments
. Since the 1996 annual repon, no new assessments tc, the BASH large-break ECCS Evaluation Model that would affect the VEGP LBLOCA PCT analysis have been identified.
LBLOCA ECCS Evaluation Model Assessment Summaly.
The absolute sum of the LBLOCA PCT assessments since the last LBLOCA significant error report (Reference 1) is less than 507.
10 CFR 50.59 Evaluation Assessments There are three plant modifications pursuant to 10 CFR 50.59 which affect the LBLOCA analysis results. The combined PCT effects from the two evaluations for the permanent radiation shield and for the trisodium phosphatr baskets result in only a IT PCT assessment.
A third plant modification, the addition of metal mass in containment, is heir.g tracked for completeness, even though the PCT penalty is much less than IT and is reponed as OT.
Licensing Basis _LBLOCA PCT
- Based on the above discussions concerning the VEGP-specific application of the Westinghouse
. BASH large-break ECCS Evaluation Model, the licensing basis LBLOCA PCT is as follows:
A.'
1997 Annual Report LBLOCA BASH ECCS Model Analysis-of-Racord l
- 1. LOCBART Analysis Result 1915'F*
- 2. LOCBART Clad Creep and Burst Error 4*F *
- 3. Evaluation for Containment Purging
+ 10*F
- 4. Evaluation for +/- 6*F Uncenainty Band
. + 117
- 5. Evaluation for Transition Cycle Penalty (Unit 1 only)
+ 50T
- 6. 100 psig Backfill Pressure with ZlRLO Clad
+ 21T
- 7. ZIRLO Clad Fuel Rods
+
5F B.
Prior BASH Large-Break ECCS Model Assessments j
Steam Generator Flow Area Application, Structural Metal Heat Modeling, LUCIFER Error Corrections, and Translation of Fluid Conditions from SATAN to LOCTA as reported to the NRC in Reference 1.
6T C.
1997 BASH Large-Break ECCS Model Assessment No new assessments were identified in 1997.
1 ATTACHMENT Page 4
~D.
10 CFR 50.59 Evaluations
- 1. Permanent Radiation ShieldffSP Baskets
+ 1*F
=
(Unit 2) 1911*F 9
=
- The 1996 annual report combined the 19157 LOCBART analysis result and the -47 LOCBART clad creep and burst error into a single value of 1911*F. These values are now being reported separately.
Conclusion When the effects of assessments to the B ASH ECCS Evaluation Model and of safety evaluations were combined with the VEGP LBLOCA Analysis results, it was determined that compliance with the requirements of 10 CFR 50.46 is being maintained for both Units 1 and 2.
SMALL-BREAK LOCA EC_CS Rvaluation Model Sinct the last annual report (Reference 1), no new assessments were identified against the small-break LOCA (SBLOCA) analysis PCT for VEGP Units 1 and 2. The current SBLOCA analysis results are based on the earlier Westinghouse NOTRUMP small-break ECCS Evaluation Model (Reference 6) as approved by the NRC for VEGP-specific application (Refe ences 4 and 5) and the latest acceptable SBLOCTA model. The limiting size break analysis continues to assume the follow'mg information imponant to the SBLOCA analyses:
o 17x17 VANTAGE-5 Fuel Assembly o
Core Power = 1.02
- 3565 MWT o
Vessel Average Temperature = 571.9T o
Steam Generator Plugging Level = 10%
o Fq = 2.48 at 9.5 fl o
FAH = 1.70 r
ATTACHMENT-Page 5 For VEGP Units I and 2, the limiting size small-break continues to be a three-inch equivalent diameter break in the cold leg. The SBLOCA analysis-of-record SBLOCTA calculated PCT t value is 17707 as discussed below..
In the 1996 annual report (Reference 1), Southern Nuclear reported an SBLOCTA Reanalysis Result PCT value of 1778*F, which reflected a +87 change from the previously reported value of 1770T (Reference 9). The change of+87 was for corrections to an SBLOCTA fuel rod initialization error. Southern Nuclear has decided to track the SBLOCTA fuel rod initialization error as a separate item under Analysis-of-Record rather than combining it with the SBLOCTA Analysis Result. Therefore, the SBLOCTA Analysis Result is reported as 17707, and the Analysis-of-Record category includes a separate +8T for the SBLOCTA fuel rod initialization error.
The steam generator lower level tap relocation and T,v, uncertainty items continue to be listed
- separately per the format of WCAP-13451. The items are listed separately because these items are not explicitly modeled. The PCT assessment values on these items are 157 and 47, respectively. A PCT assessment of 30T is also listed separately for Burst and Blockage / Time in Life.
VEGP began using ZIRLO clad fuel rods in the Unit 2 Cycle 6 and Unit 1 Cycle 8 cores. ~ The use ofZlRLO clad fhel rods results in a penalty of 3*F PCT as calculated in the latest acceptable SBLOCTA model. This penalty applies to both IFBA and non-IFBA rods.
Prior NOTRUMP Small-Break ECCS Evaluation Model Assessments In the following table and as reported in References 1 and 7, five prior model assessments have been combined. These assessments are: (1) Safety Injection (SI) Flow into the Broken RCS Loop / Improved Steam Condensation Model, (2) Drift Flux Flow Regime Error, (3) LUCIFER
- Error Corrections, (4) Boiling Heat Transfer Correlation Error, and (5) Steam Line Isolation Logic Error.
The NOTRUMP specific enthalpy error continues to be listed separately in accordance with WCAP-13451 since it was not combined with the prior model assessments (see Reference 1).
1997 NOTRUMP Small-Break ECCS Evaluation Mcdel As_sgsgrnenJs No new assessments have been identified to the NOTRUMP SBLOCA ECCS Evaluation Model that would affect the VEGP analysis results.
A'ITACHMENT Page 6 SBLOCA ECCS Model Assessment Summary The absolute sum of the new SBLOCA PCT assessments since the last SBLOCA significant error report, Reference 8, is less than 507 for the VEGP NOTRUMP SBLOCA ECCS model.
10 CFR 50.59 Evaluation Assessments There are three plant modifications pursuant to 10 CFR 50.59 which aff'ect the SBLOCA analysis results for VEGP Unit 1. These are: (1) annular pellet blankets, (2) additional metal mass in containment, and (3) loose part in the RCS. The PCT penalty on additional metal mass in containment is the only one of the three which is applicable to VEGP Unit 2. The PCT penalty on additional metal mass in containment is being tracked for completeness, even though the PCT penalty is much less than IT and is reported as 07.
Licensing Basis SBLOCA PCT Based on the above discussions concerning the VEGP-specific application of the Westinghouse NOTRUMP small-break ECCS Evaluation Model, the licensing basis SBLOCA PCT is as follows:
A.1997 Annual Report SBLOCA NOTRUMP ECCS Model Analysis-of-Record
- 1. SBLOCTA Analysis Result 1770T*
- 2. SBLOCTA Fuel Rod Initialization Error
+
8T*
- 3. Evaluation for Steam Generator Lower Level Tap Relocation
+ 15 7
- 4. Evaluation for +/- 6*F Uncertainty Band
+
47
- 5. Burst and Blockagefrime in Life
+ 30 F
- 6. ZIRLO Clad Fuel Rods
+
3F B.. Prior NOTRUMP Small-Break ECCS Model Assessments
- 1. Safety Injection Flow into Broken RCS Loop / Improved 17T Steam Condensation Model, Drift Flux Flow Regime, LUCIFER Error Corrections, Boiling Heat Transfer Correlation Error, and Steam Line Isolation Logic Error as reponed to the NRC in Reference 7.
- 2. NOTRUMP Specific Enthalpy Error
+ 20T C.1997 NOTRUMP Small-Break ECCS Model Assessments No new assessments were identified in 1997.
ATTACHMENT Page 7 D.10 CFR 50.59 Evaluations
- 1. Annular Pellet Blankets (VEGP Unit I only)
+ 10 7
- 2. Addition ofMetal Mass in Containment 07
- 3. Loose Part (VEGP Unit 1 only)
+
2*F -
Licensing Basis SBLOCA PCT (Unit 1) =
.lMi F (Unit 2) =
1131'F -
- The 1996 annual report combined the 1770*F SBLOCTA analysis result and the +87 SBLOCTA fuel rod initialization error into a single value of 1778T. These values are now being reported separately, i
Conclusion When the effects of assessments to the NOTRUMP ECCS Evaluation Model and the effects of safety evaluations were combined with the VEGP SBLOCA analysis results, it was determined that compliance with the requirements of 10 CFR 50,46 is being maintained for both Units 1 and 2.
REFERENCES 1.
LCV-0998, "Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Evaluation Models 1996 Annual Report and Significant Error Report," letter from C. K. McCoy (GPC) to USNRC, dated March 31,1997.
2.
WCAP-13451, " Westinghouse Methodology for implementation of 10 CFR 50.46 Reporting," dated October 1992.
3.
"The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A, Rev. 2,. (Proprietary), March 1987.
4.
Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Nos. 43 and 44 to Facility Operating License NPF-68 and Amendment Nos. 23 and 24 to Facility Ooerating License NPF-81, attachment to letter from Hood (USNRC) to Hairston (GPC), dated September 19,1991.
4TrACHMENT iage8 5.
Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendm_.sg No. 60 to Facility Operating License NPF-68 and Amendment No. 39 to Facility Operatina License NPF-81. attachment to letter from Hood (USNRC) to Hairston (GPC),
dated March 22,1993.
6.
" Westinghouse Small-Break ECCS Evaluation Model Using the NOTRUMP Code,"
WCAP-10054-P-A (Proprietary) and WCAP-10081-A (Non-Proprietary), August 1985.
7.
LCV-0579, "Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Evaluation Models 1994 Annual Report," letter from C. K. McCoy (GPC) to USNRC, dated March 17, 1995.
8.
LCV-0327-B,"Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Evaluation Models Significant Change Report," letter from C. K. McCoy (GPC) to USNRC, dated December 8,1994.
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l 9.
LCV-0780,"Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Evaluation Models 1995 Annual Report," letter from C. K. McCoy (GPC) to USNRC, dated March 25, 1096.-
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