ML20206D588
| ML20206D588 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 04/29/1999 |
| From: | Beasley J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCV-1342, NUDOCS 9905040160 | |
| Download: ML20206D588 (10) | |
Text
_
J. Barnie Beesley, Jr., P.E.
Soutt:ern Nuclear Vice President Operating Company,Inc.
Vogt!e Project 40 invemass Center Pekway PO Box 1295 Birmingham. Alabama 35201 Tel 205 932 7110 Fax 205 992 0403 SOUTHERN COMPANY Energy ro Serve l' ur World
- o April 29, 1999 LCV-1342 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Ladies and Gentlemen:
VOGTLE ELECTRIC GENERATING PLANT 10CFR50.46 ECCS EVALUATION MODELS 1998 ANNUAL REPORT Attached is the Southern Nuclear Operating Company (Southern Nuclear) 10CFR50.46 Emergency Core Cooling System (ECCS) Evaluation Models 1998 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 changes and errors assessed against the Vogtle Electric Generating Plant (VEGP) ECCS Evaluation Models since the 1997 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 1998 Annual Report, compliance with the PCT requirements of 10CFR50.46 continues to be maintained when the efTects of plant design changes are combined with the effects of the ECCS Evaluation Models assessments applicable to VEGP Units 1 and 2.
1 Ifyou have any questions regarding this report, please contact this office.
i Sincere,
by hD
" *^ Y' na D
Attachment 9905040160 990429 ADOCK0500g4 DR
U. S. Nuclear Regulatory Commission Page 2 JBB/RJF xc: Southern Nuclear Operating Comoany Mr. J. T. Gasser Mr. M. Sheibani
. SNC Document Management U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. R. Assa, Licensing Project Manager, NRR Mr. J. Zeiles, Senior Resident Inspector, Vogtle LCV-1342 l
1 i
I ATTACIIMENT VOGTLE ELECTRIC GENERATING PLANT 10,CFR 50.46 ECCS EVALUATION.MODELS 1998 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 of discovery 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 clad temperature (PCT) different by more than 50T from the temperature calculated for the limiting transient using the last acceptable model, or a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 507.
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 1997 annual report (Reference 1). This annual report has been prepared in accordance with the methodology presented in WCAP-13451 (Reference 3). The LBLOCA and SBLOCA analyses, Evaluation Model assessments, and 1
safety evaluation results reported herein will be included in the next VEGP Final Safety Analysis Report (FSAR) update.
LARGE-BREAK LOCA ECCS Evaluation Model Since the previous report (Reference 1), one new assessment against the VEGP LBLOCA analysis has been identified. A 43 T PCT penalty for increased accumulator line resistance has been assessed. This 43 7 PCT penalty is the only LBLOCA assessment since the last VEGP LBLOCA significant error report (Reference 2). Therefore, the absolute sum of the LBLOCA PCT assessments since the last significant error report remains below 50 7.
The LBLOCA analysis results are based on the Westinghouse BASH large-break ECCS Evaluation Model (Reference 4), as approved by the NRC for VEGP-specific application (References 5 and 6) 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 = 571.9T o
Steam Generator Plugging Level = 10%
o Fq = 2.50 o
FAH = 1.65
F 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 1915*F.
The Analysis-of-Record category continues to include an assessment of-4 'F for the LOCBART clad creep and burst error.
The containment purge, T, uncertainty, and transition core penalty items continue to be listed separately. The items are listed separately because these items are not explicitly modeled. The PCT assessment values on these items remain 10,11, and 50*F, respectively. The VEGP Unit 2 core contained all VANTAGE-5 fuel during 1998. Therefore, the Transition Cycle Penalty did not apply to VEGP Unit 2.
VEGP used ZIRLO clad fuel rods in the Unit I and Unit 2 cores during 1998. The use of ZIRLO clad fuel rods results in a penalty of 5*F PCT as calculated by the latest acceptable LOCBART model.
VEGP used ZIRLO clad IFBA fuel rods with a backfill pressure of 100 psig in the Unit I and
. Unit 2 cores in 1998. The use of ZIRLO clad IFBA rods results in a penalty of 21*F PCT as calculated by the latest acceptable LOCB ART model.
VEGP cores continue to contain a mix of Zircaloy and ZIRLO clad fuel rods and, IFB A and non-IFBA rods. VEGP will continue to show an analysis-of-record LOCBART calculated PCT value based on non-IFBA, Zircaloy fuel rods (1915'F), and will apply PCT penalties to address the use ofZIRLO clad fuel rod.
Prior 10CFR50.46 BASH Large-Break ECCS Evaluation Model Assessments In the following table and as reported in Reference 2, four prior model assessments have been combined into a single assessment of-6 F. These assessments are: (1) Steam Generator Flow Area Application, (2) Structural Metal Heat Modeling, (3) LUCIFER Error Correction, and
)
(4) Translation of Fluid Conditions from SATAN to LOCTA.
1
ATTACHMENT Page 3 199810CFR50.46 BASH Large-Break ECCS Evaluation Model Assessments Since the 1997 annual report, one new assessment to the B ASH large-break ECCS Evaluation Model that would affect the VEGP LBLOCA PCT analysis has been identified. A 43 T PCT penalty for an increase in accumulator line resistances has been assessed.
The ECCS accumulator check valves are tested as part of the Inservice Test Program. As part of the test, the accumulator line resistances are verified for comparison with the safety analyses. This is done as a backup to the check valve test. During measurement of the accumulator line resistances in the Unit 2 end-of-cycle 6 outage (Spring 1998), it was discovered that the measurements were not consistent with the safety analyses. The LOCA analyses were revised to reflect higher line resistances. The increased line resistances in the analyses resulted in an increase in the peak clad temperature (PCT) for the limiting break.
LBLOCA 10CFR50.46 ECCS Evaluation Model Assessment Summary The absolute sum of the LBLOCA PCT assessments since the last LBLOCA significant error report (Reference 2) is less than 50*F.
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 phosphate baskets result in only a IT PCT assessment.
A third plant modification, the addition of metal mass in the containment, is being tracked for completeness, even though the PCT penalty is much less than IT and is reported as 07.
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.
1998 Annual Report LBLOCA BASH ECCS Model Analysis-of-Record
- 1. LOCBART Analysis Result 1915T
- 2. LOCBART Clad Creep and Burst Error 47
- 3. Evaluation for Containment Purging
+ 10T
- 4. Evaluation for +/- 6T Uncertainty Band
+ 1IT
- 5. Evaluation for Transition Cycle Penalty (Unit 1 only)
+ 507
- 6. 100 psig Backfill Pressure with ZIRLO Clad
+ 21T
- 7. ZIRLO Clad Fuel Rods
+
'l j
l ATTACHMENT Page 4 3
B.
Prior 10CFR50.46 BASH Large-Break ECCS Model Assessments l
Steam Generator Flow Area Application, Structural Metal Heat Modeling, LUCIFER Error Corrections, and Translation i
of Fluid Conditions from SATAN to LOCTA as reported to the NRC in Reference 2.
6*F C,
199810CFR50.46 BASH Large-Break ECCS Model Assessment Increased Accursuiu. Line Resistances
+ 437 D.
10 CFR 50.5P Evalud.cas
- 1. Permanent Radiation Shield / TSP Baskets
+
=
(Unit 2) 19967
=
Conclusion When the effects of assessments to the BASH 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 VEGP Units 1 and 2.
SMALL-BREAK LOCA ECCS Evaluation Model Since 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 7) as approved by the NRC for VEGP-specific application (References 5 and 6) and the latest acceptable SBLOCTA model. The limiting size break analysis continues to assume the following information important 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 = 2.58 o
FAH = 1.70
ATTACHMENT Page 5 For VEGP Units 1 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 value is 1770*F.
The Analysis-of-Record category continues to include an assessment of +8 *F for the SBLOCA fuel rod initialization error.
i The steam generator lower level tap relocation and T,v, uncertainty items continue to be listed separately. The items are listed separately because these items are not explicitly modeled. The PCT assessment values on these items are 15*F and 4*F, respectively. A PCT assessment of 30*F is also listed separately for Burst and Blockage / Time in Life.
VEGP used ZIRLO clad fuel rods in Unit I and Unit 2 during 1998. The use of ZIRLO clad fuel 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 10CFR50.46 NOTRUMP Small-Break ECCS Evaluation Model Assessments In the following table and as reported in References 2 and 8, 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) Drill 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 10).
199810CFR50.46 NOTRUMP Small-Break ECCS Evaluation Model Assessments No new assessments have been identified to the NOTRUMP SBLOCA ECCS Evaluation Model that would affect the VEGP analysis results.
=
ATTACHMENT Page 6 s
T I
SBLOCA 10CFR50.46fCCS Model Assessment Summarv' The absolute sum of the new SBLOCA PCT assessments since the last SBLOCA significant error report, Reference 9, is less than 50 F for the VEGP NOTRUMP SBLOCA ECCS model.
i 10 CFR 50.59 Evaluation Assessments i
There are two plant modifications pursuant to 10 CFR 50.59 which affect the SBLOCA analysis results for VEGP Unit 1. These are: (1) annular pellet blankets, and (2) loose part in the RCS. The PCT penalty on annular pellet blankets is the only one of the two which is applicable to VEGP Unit 2. The line item on additional metal mass in the containment, which was included in the 1997 annual report (Reference 1) as a 0 "F penalty, has been deleted because the SBLOCA results are insensitive to containment pressure, and containment pressure is not modeled in NOTRUMP.
l 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.1998 Annual Report SBLOCA NOTRUMP ECCS Model Analysis-of-Record
- 1. SBLOCTA Analysis Result 1770*F
- 2. SBLOCTA Fuel Rod Initialization Error
+
8"F
- 3. Evaluation for Steam Generator Lower Level Tap Relocation
+ 15*F
- 4. Evaluation for +/- 6*F Uncertainty Band
+
4*F
- 5. Burst and Blockage / Time in Life
+ 30 F
- 6. ZIRLO Clad Fuel Rods
+
3F B. Prior 10CFR50.46 NOTRUMP Small-Break ECCS Model Assessments i
- 1. Safety Injection Flow into Broken RCS Loop / Improved 17"F Steam Condensation Model, Drift Flux Flow Regime, LUCIFER Error Corrections, Boiling Heat Transfer Correlation Error, and Steam Line Isolation Logic Error as reported to the NRC in Reference 8.
- 2. NOTRUMP Specific Enthalpy Error
+ 20 F C.199810CFR50.46 NOTRUMP Small-Break ECCS Model Assessments No new assessments were identified in 1998.
0*F
l ATTACHMENT i
Page 7 D.10 CFR 50.59 Evaluations
- 1. Annular Pellet Blankets
+ 10*F
- 2. Loose Part (VEGP Unit 1 only)
+
2*F Licensing Basis SBLOCA PCT (Unit 1) =
1845 F (Unit 2) =
1843 F 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 VEGP Units 1 and 2.
REFERENCES 1.
LCV-1190,"Vogtle Electric Generating Plant,10CFR50.46 ECCS Evaluation Models 1997 Annual Repcrt," letter from C. K. McCoy (SNC) to USNRC, dated March 30, 1998.
2.
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 (SNC) to USNRC, dated March 31,1997.
3.
WCAP-13451, " Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," dated October 1992.
4.
"The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A, Rev. 2, (Proprietary), March 1987.
5.
Safety Evaluation by the Office ofNuclear 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.
l l
ATTACHMENT Page 8 6.
Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment j
No. 60 to Facility Operating License NPF-68 and Amendment No. 39 to Facility Operating License NPF-81, attachment to letter from Hood (USNRC) to Hairston (GPC),
j dated March 22,1993.
7.
" Westinghouse Small-Break ECCS Evaluation Model Using the NOTRUMP Code,"
WCAP-10054-P-A (Proprietary) and WCAP-10081-A (Non-Proprietary), August 1985.
8.
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.
9.
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.