NL-04-1042, Plan Joseph M. Farley Nuclear Plant, and Vogtle Electric Generating Plant, 10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2003

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Plan Joseph M. Farley Nuclear Plant, and Vogtle Electric Generating Plant, 10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2003
ML041830368
Person / Time
Site: Hatch, Vogtle, Farley  Southern Nuclear icon.png
Issue date: 06/29/2004
From: Stinson L
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-1042
Download: ML041830368 (28)
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Text

L M. Stinson (Mike) Southern Nuclear Vice President Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.5181 Fax 205.992.0341 A SOUTHERN June 29, 2004 COMPANY Energy to Serve Your Wo rld' Docket Nos.: 50-321 50-348 50-424 NL-04-1042 50-366 50-364 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Joseph M. Farley Nuclear Plant Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2003 Ladies and Gentlemen:

Pursuant to the reporting requirements of 10 CFR 50.46 (aX3)(ii), Southern Nuclear Operating Company (SNC) is submitting the emergency core cooling system (ECCS) evaluation model annual reports for Hatch Nuclear Plant Units 1 and 2, Farley Nuclear Plant Units 1 and 2, and Vogtle Electric Generating Plant Units 1 and 2.

These annual reports summarize the nature of and estimated effect of any changes or errors in the ECCS models for the period from January 1, 2003 through December 31, 2003.

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely, L. M. Stinson LMSIRJF/daj

Enclosures:

1. Edwin I. Hatch Nuclear Plant 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003
2. Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003
3. Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 OD

U. S. Nuclear Regulatory Commission NL-04-1042 Page 2 cc: Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. L. M. Stinson, Vice President, Plant Farley Mr. H. L. Sumner, Jr., Vice President, Plant Hatch Mr. J. T. Gasser, Vice President, Plant Vogtle Mr. D. E. Grissette, General Manager - Plant Farley Mr. G. R. Frederick, General Manager - Plant Hatch Mr. W. F. Kitchens, General Manager - Plant Vogtle RType: CFA04.054; CHAO2.004; CVC7000; LC# 14064 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. S. E. Peters, NRR Project Manager - Farley Mr. C. Gratton, NRR Project Manager - Hatch Mr. C. Gratton, NRR Project Manager - Vogtle Mr. C. A. Patterson, Senior Resident Inspector - Farley Mr. D. S. Simpkins, Senior Resident Inspector - Hatch Mr. J. Zeiler, Senior Resident Inspector - Vogtle

Enclosure 1 Edwin L Hatch Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003

Enclosure 1 Edwin I. Hatch Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 BACKGROUND In accordance with 10 CFR 50.46(aX3Xii), this annual report summarizes the nature of and estimated effect of any changes or errors in the emergency core cooling system (ECCS) model for the period from January 1, 2003 through December 31, 2003 for Hatch Nuclear Plant Units 1 and 2.

DISCUSSION Updated limiting licensing basis peak clad temperatures (PCTs) applicable to Hatch are provided in the following table.

In 2003 Hatch Units 1 and 2 operated with both GE13 and GE14 fuel in their cores. Therefore, the updated licensing basis PCTs are provided for both GE13 and GE14 fuel. The following table begins by listing the baseline ECCS-LOCA evaluations for GE13 fuel (Reference 1)and GE14 fuel (Reference 2).

The next section of the table lists the applicable changes or errors and their estimated effect on PCT that have previously been reported to the NRC (References 3, 4, and 5).

The final section of the table lists those applicable changes or errors and their estimated effect on PCT which SNC has been notified of by GE during the period from January 1, 2003 through December 31, 2003. In addition, the impact on PCT is reported of a planned 2004 SNC plant change to increase nominal reactor dome pressure by 10 psi. Additional information on the three applicable changes or errors identified during 2003 follows.

Impact of SAFER Level/Volume Table Error on the PCT In GE 10 CFR 50.46 Notification Letter 2003-01 (Reference 6), GE reported that an error was found in the initial level/volume table for SAFER; specifically, the value of the initial water level differed from the value used in the original reactor pressure vessel (RPV) level/volume calculation.

This error resulted in an incorrect volume split in the nodes above and below the water surface, and incorrect initial liquid mass. The corrected liquid volume is higher if the water level has been lowered relative to the original RPV volume calculation. Higher initial liquid volume results in lower PCT. GE estimated the effect on PCT for Hatch to be 0 0 F for GE13 and -5 IF for GE14.

Impact of SAFER Initial Separator Pressure Drop Error on the PCT In GE 10 CFR 50.46 Notification Letter 2003-03 (Reference 7), GE reported an error in the initial steam separator pressure drop used as an input to the SAFER model. Specifically, either the incorrect product line dependent loss coefficient was applied or a term to account for the hydrostatic pressure was erroneously included (this term is calculated separately by the SAFER model). These errors resulted in a higher initial steam separator pressure drop and overly restricted the flow through the separator during the LOCA event. The corrected initial steam separator value resulted in higher flow through the core during the LOCA event. This in turn resulted in higher El-l 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 core water levels and a small decrease in PCT for jet pump plants. GE conservatively estimated the effect on PCT for Hatch to be 0 IF for GE13 and N/A for GE14.

Impact of 10-PSI Nominal Reactor Dome Pressure Increase on the PCT In support of implementation in 2004 of a 10 psi nominal reactor dome pressure increase, ECCS performance with both GE13 and GE14 fuel was evaluated (Reference 8). The effect on GE14 fuel of the 10 psi nominal reactor dome pressure increase was already considered and reflected in the previously discussed baseline ECCS-LOCA evaluation for GE14 fuel (Reference 2). GE conservatively estimated the effect of the 10 psi nominal reactor dome pressure increase on PCT to be 10 F for GE13.

CONCLUSION As documented in the following table, the updated Hatch limiting licensing basis PCTs for GE13 and GE14 remain in compliance with 10 CFR 50.46(b)(1), specifically requiring that the limiting licensing basis PCT shall not exceed 2200 IF. As such, there is no need for reanalysis or taking any other actions in accordance with 10 CFR 50.46(aX3Xii) because compliance with 10 CFR 50.46(b)(1) has been maintained.

Report Description of Estimated PCT Change ( IF) Updated PCT Period Change or Errorin ECCS GE13 Fuel GE14 Fuel GE13 GE14 Evaluation PCT Absolute PCT Absolute Fuel Fuel Change Value Change Value SAFER/

GESTR-LOCA N/A N/A N/A N/A 1688 N/A Analysis dated Baseline March, 1997 Evaluations (Ref. 1)

SAFER/

GESTR-LOCA N/A N/A N/A N/A N/A 1820 Analysis dated March, 2002 (Ref. 2)

E1-2

' Enclosure 1 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 Report Description of Estimated PCT Change (IF) Updated PCT Period Change or Error in ECCS GE13 Fuel GE14 Fuel GE13 GE14 Evaluation PCT Absolute PCT Absolute Fuel Fuel Change Value Change Value Hatch 50.46 Annual Report 10 20 N/A N/A 1698 N/A for 2000 (Ref. 3)

Previously Hatch 50.46 Reported 30 Day Report 100 100 N/A N/A 1798 N/A Changes or dated 5/21/01 Errors (Ref. 4)

Hatch 50.46 Annual Report 15 15 0 0 1813 1820 for 2002 (Ref. 5)

GE 50.46 Notification 0 0 5 5 1813 1815 Letter 2003-01 00- 83 11 dated 5/6/03 2003 (Ref. 6)

Changes or GE 50.46 Errors Notification 00NA NA 11 /

Letter 2003-03 0 0 N/A N/A 1813 N/A dated 5/6/03 (Ref. 7) 10-PSI Dome Pressure Increase Evaluation 10 10 N/A N/A 1823 N/A dated July, 2003 (Ref. 8)

E1-3 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 REFERENCES

1. NEDC-32720P, "Hatch Units 1 and 2 SAFER/GESTR Loss-of-Coolant Accident Analysis,"

dated March 1997.

2. GE-NE-0000-0000-9200-02P, "Hatch Units 1 and 2 ECCS-LOCA Evaluation for GE14,"

dated March 2002.

3. SNC Letter HL-6028, H. L. Sumner, Jr. to NRC, "Reporting of Changes and Errors in ECCS Evaluation Models," dated January 31, 2001.
4. SNC Letter HL-6090, H. L. Sumner, Jr. to NRC, "Reporting of Changes and Errors in ECCS Evaluation Models," dated May 21, 2001.
5. SNC Letter NL-03-0999, J. B. Beasley, Jr. to NRC, "10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2002," dated June 2, 2003.
6. E-mail from Margaret E. Harding (Global Nuclear Fuel (GNF)) to Ken S. Folk (SNC), "10 CFR 50.46 Notification - 2003 Southern Co.," dated May 16, 2003.
7. E-mail from Margaret E. Harding (GNF) to Ken S. Folk (SNC), "10 CFR 50.46 Notification -

2003 Southern Co.," dated May 16, 2003.

8. GE-NE-0000-0003-0634-01, Revision 1,"Edwin I. Hatch Nuclear Plant, Units 1 and 2, 10-PSI Dome Pressure Increase," dated July 2003.

E1-4

Enclosure 2 Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003

Enclosure 2 Joseph M. Farley Nuclear Plant 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 BACKGROUND In accordance with 10 CFR 50.46(a)(3)(ii), this annual report summarizes the nature of and estimated effect of any changes or errors in the emergency core cooling system (ECCS) model for the period from January 1, 2003 through December 31, 2003 for Farley Nuclear Plant Units 1 and 2.

DISCUSSION In Reference 1, information was submitted to the NRC regarding modifications to the Westinghouse large-break and small-break Loss-of-Coolant Accident (LOCA) ECCS Evaluation Models as applicable to the Farley analyses for the calendar year 2002.

The following presents an assessment of the effects of modifications to the Westinghouse ECCS Evaluation Models on the Farley LOCA analysis results since the 2002 annual report (Reference 1) for the calendar year 2003. This annual report has been prepared in accordance with the Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting (WCAP-13451, Reference 2), with the exception of plant changes. Starting in 2001, a change in the Westinghouse reporting methodology was made to include the 50.59 Plant Change PCT values as a part of the 50 IF error reporting section. The 2003 annual report (contained herein) is consistent with the change implemented in the 2001 annual report.

Unit 2 implemented the Reactor Internals Upflow Conversion Program (Reference 3) in 2002, and as such a new PCT rack-up reflecting the new upflow configuration analysis is presented here for Unit 2.

Large-Break LOCA Table IA shows the LBLOCA PCT rack-ups for both Unit 1 and Unit 2 for Reflood 1 (Reference 4). Table 1B shows the corresponding large-break LOCA PCT rack-ups for Reflood 2 (Reference 4).

LBLOCA ECCS MODEL ANALYSIS-OF-RECORD The large-break LOCA analyses for Farley Units 1 and 2 were examined to assess the effects of the changes and errors in the Westinghouse large-break LOCA ECCS Evaluation Model on PCT results.

E2-1 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 The large-break LOCA analysis-of-record results for Farley Units 1 and 2 were calculated using Westinghouse's BE-LOCA analysis (References 1 and 4).

The Unit 1 and Unit 2 analyses assumed the following information important to the large-break LOCA in the BE-LOCA analysis (References 1 and 4). One analysis was used to bound both Farley Unit 1 and Unit 2.

Core Power = 2775 MWT 17x17 VANTAGE+ Fuel Assembly FQ = 2.50 for VANTAGE+ Fuel FAH = 1.70 for VANTAGE+ Fuel SGTP = 20%

For Farley Units 1 and 2, the limiting size break analysis-of-record is a split break of the cold leg piping with a discharge coefficient of CD = 1.0. The limiting PCT values determined for the Unit 1 and Unit 2 large break LOCAs are shown in Table IA (Reflood 1).

PRIOR LBLOCA ECCS MODEL ASSESSMENTS Prior 10 CFR 50.46 Assessments Reported as Significant There are no LBLOCA 10 CFR 50.46 prior assessments reported as significant.

Prior 10 CFR 50.59 Assessments The following two plant change assessments were reported in the last submittal (Reference 1) and occurred prior to 2001.

The addition of permanent storage boxes in containment was evaluated and found not to cause a change to PCT (Reference 6).

The finalization of Replacement Steam Generator Data was evaluated and found not to cause a change to PCT (Reference 1).

E2-2 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 CURRENT LBLOCA ECCS MODEL ASSESSMENTS The following changes and errors in the Westinghouse ECCS Evaluation Model would affect the BE-LOCA Model.

Prior 10 CFR 50.46 Reported Assessments The following two assessments were reported in the last PCT submittal (Reference 1).

Accumulator line/Pressurizer Surge Line Data It was determined that the design and actual plant accumulator line piping schedule were not the same. A Farley specific BE-LBLOCA sensitivity analysis resulted in a 41 0F benefit for the first reflood and a 9 IF benefit for the second reflood when actual plant data was modeled (Reference 7). This assessment is applicable to Unit 1 and Unit 2.

Decay Heat Uncertainty error in Monte Carlo Calculation It was determined that an error existed in the calculation of decay heat uncertainty in the Monte Carlo calculation of the 95th percentile PCT for BE-LBLOCA (Reference 9). This caused an 8 IF penalty for Unit 1 and 2 on Reflood I only.

2003 10 CFR 50.46 PCT Assessments None.

CURRENT PLANNED PLANT CHANGE EVALUATIONS Starting with the 2001 annual report (Reference 1), the 10 CFR 50.59 Plant Change PCT values have been considered to be a part of the 50 0F error reporting section. The 2003 annual report (contained herein) is consistent with the changes implemented in the 2001 annual report.

Prior 10 CFR 50.59 Model Assessments None.

2003 Planned Plant Changes None.

E2-3 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 TOTAL RESULTANT LBLOCA PCT As discussed above, the changes and errors to the Westinghouse large-break LOCA ECCS Evaluation Model could affect the large-break LOCA analysis results by altering the PCT. As shown in Table IA and Table IB, the large-break LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 2200 IF.

E24 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 Small-Break LOCA Table 2 shows the small-break LOCA PCT rack-ups for both Unit 1 and Unit 2.

SBLOCA ECCS MODEL ANALYSIS-OF-RECORD The small-break LOCA analyses for Farley Units 1 and 2 were also examined to assess the effects of the changes and errors to the Westinghouse small-break LOCA ECCS Evaluation Models on PCT results. The small-break LOCA ECCS analysis results were calculated using the NOTRUMP small-break LOCA ECCS Evaluation Model (Reference 5). As noted earlier, the Unit 2 re-analysis reflects the Reactor Internals Upflow Conversion implemented in 2002 (Reference 3).

The Unit 1 and Unit 2 analyses assumed the following information important to the small-break LOCA analyses:

Unit 1 Unit 2 Core Power = 1.02 X 2775 MWT Core Power = 1.02 x 2775 MWT 17x17 VANTAGE+ Fuel Assembly 17x17 VANTAGE+ Fuel Assembly FQ = 2.50 FQ = 2.50 FAH = 1.70 FAH = 1.70 Upflow Configuration Upflow Configuration For Farley Units I and 2, the limiting size break analysis-of-record for the VANTAGE+ fuel analysis is a 3-inch diameter break in the cold leg. The limiting PCT values determined for the Unit 1 and Unit 2 17x17 VANTAGE+ small-break are shown in Table 2.

PRIOR SLBLOCA ECCS MODEL ASSESSMENTS Prior 10 CFR 50.46 Assessments Reported as Significant The following SBLOCA 10 CFR 50.46 assessment was reported in March 2000 as significant.

An overall PCT benefit of 62 'F for Unit 1 for the "Burst and Blockage/Time in Life" penalty resulted from the SPIKE computer code correlation revision. (Reference 11).

E2-5 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 Prior 10 CFR 50.59 Assessments The following three plant change assessments were reported in the last submittal (Reference 1) and occurred prior to 2001.

The addition of permanent storage boxes in containment was evaluated and found not to cause a change to PCT (Reference 6).

The finalization of Replacement Steam Generator Data resulted in a 62 IF benefit for Unit 1 (Reference 10).

Annular pellets were determined to have a 10 IF penalty for SBLOCA results for Unit 1 (Reference 8).

Note that the Unit 2 result (in Table 2) is unaffected by these prior 50.59 plant changes. The reason is that the Unit 2 Upflow Conversion implemented in 2002 required a small-break LOCA re-analysis that included the above changes explicitly.

CURRENT SBLOCA ECCS MODEL ASSESSMENTS The following changes and errors were identified:

Prior 10 CFR 50.46 Reported Assessments The following assessments were reported in the last PCT submittal (Reference 1).

NOTRUMP Mixture Level Tracking/Region Depletion Errors Several closely related errors have been discovered in how NOTRUMP deals with the stack mixture level transition across a node boundary in a stack of fluid nodes. As previously reported, the impact of this revision on the SBLOCA results has been determined to be a 13 IF penalty for Unit 1. In addition, the associated change in Burst and Blockage/Time in Life Components was an additional 12 IF for Unit 1. Thus, the total change was 25 IF for Unit 1. This error does not impact Unit 2's re-analysis result (see previously discussed Reactor Internals Upflow Conversion),

since the re-analysis was performed with the corrected version of NOTRUMP.

2003 10 CFR 50.46 PCT Assessments None.

E2-6 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 CURRENT PLANNED PLANT CHANGE EVALUATIONS Starting with the 2001 annual report (Reference 1), the 10 CFR 50.59 Plant Change PCT values have been considered to be a part of the 50 IF error reporting section. The 2003 annual report (contained herein) is consistent with the change implemented in the 2001 annual report.

Prior 10 CFR 50.59 Model Assessments None.

2003 Planned Plant Changes None.

TOTAL RESULTANT SBLOCA PCT As discussed above, the changes and errors in the Westinghouse small-break LOCA ECCS Evaluation Model could affect the small-break LOCA analysis results by altering the PCT. As shown in Table 2, the small-break LOCA analysis PCT results for both units are below the 10 CFR 50.46 limit of 2200 OF.

CONCLUSION As documented in the following tables, the updated Farley large-break and small-break LOCA analyses PCTs remain in compliance with 10 CFR 50.46(bXl), specifically requiring that the PCT shall not exceed 2200 OF. As such, there is no need for reanalysis or taking any other actions in accordance with 10 CFR 50.46(a)(3)(ii) because compliance with 10 CFR 50.46(bXl) has been maintained.

E2-7 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 REFERENCES

1. Letter from J. B. Beasley, Jr. to USNRC (NL-03-0999), "Edwin I Hatch Nuclear Plant, Joseph M. Farley Nuclear Plant, Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2002," June 2, 2003.
2. WCAP- 13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"

October 1992.

3. ALA-02-039, "Transmittal of Reactor Internals Upflow Conversion Program Engineering Report, J. M. Farley Nuclear Plant Unit 2," June 2002 (also see WCAP-1 5974, November 2002).
4. ALA-04-28, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 25, 2004.
5. "Westinghouse Small-break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), Lee, N., et. al, August 1985.
6. SECL-97-062. Rev. 1, "Effects on LOCA PCT of Adding Permanent Storage Boxes and Lead Blankets Inside Containment," October 17, 1997.
7. ALA-00-037, "Final 10 CFR 50.46 Annual Notification and Reporting," March 8, 2000.
8. WCAP-15098, "Joseph M. Farley Nuclear Plant Units I and 2 RSG Program NSSS Licensing Report," November 1998.
9. ALA-01-008, "10 CFR 50.46 Annual Notification and Reporting for 2000," March 6, 2001.
10. ALA-0I-01, "Southern Nuclear Operating Company, Joseph M. Farley Nuclear Plant Units 1 and 2, LBLOCA and SBLOCA Impacts Due to Final RSG Data for SGRP," February 11, 2000.
11. Letter from D. N. Morey to USNRC (NEL-00-0080), "Joseph M. Farley Nuclear Plant 10 CFR 50.46 Annual ECCS Evaluation Model Changes Report for 1999 and Significant Error Reports," March 29, 2000.

E2-8 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 TABLE 1A (Limitiniy for Unit 1 and Unit 2)

JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT LARGE-BREAK LOCA PCT (0 F) FOR REFLOOD 1 A. LBLOCA ECCS MODEL ANALYSIS-OF-RECORD UNIT 1 UNIT 2

1. ECCS Analysis 2056* 2056*
2. Increased Containment Spray Flow 9* 9*

Total Analysis-of-Record 2065* 2065*

B. PRIOR LBLOCA ECCS MODEL ASSESSMENTS

1. Prior 10 CFR 50.46 Assessments Reported as Significant 0 0
2. Prior 10 CFR 50.59 Assessments A. Addition of Permanent Storage Boxes in Containment 0 0 B. Finalization of Replacement Stearn Generator Data 0 0 Sum of Prior Assessments 0 0 C. CURRENT LBLOCA ECCS MODEL ASSESSMENTS
1. Accumulator Line/Pressurizer Surge Line Data -41* -41*
2. MONTECF Decay Heat Uncertainty Error 8* 8*

D. CURRENT PLANNED PLANT CHANGE EVALUATIONS

1. None 0 0 E. TOTAL RESULTANT LBLOCA PCT Total 2032* 2032*

The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points. The Analysis of Record PCT results reflect the Replacement Steam Generators analysis values.

  • See References 1 and 4 E2-9 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 TABLE lB JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT LARGE-BREAK LOCA PCT (0F) FOR REFLOOD 2 A. LBLOCA ECCS MODEL ANALYSIS-OF-RECORD UNIT 1 UNIT 2
1. ECCS Analysis 1956* 1956*
2. Increased Containment Spray Flow 1* 1*

Total Analysis-of-Record 1957* 1957*

B. PRIOR LBLOCA ECCS MODEL ASSESSMENTS

1. Prior 10 CFR 50.46 Assessments Reported as Significant 0 0
2. Prior 10 CFR 50.59 Assessments A. Addition of Permanent Storage Boxes in Containment 0 0 B. Finalization of Replacement Steam Generator Data 0 0 Sum of Prior Assessments 0 0 C. CURRENT LBLOCA ECCS MODEL ASSESSMENTS
1. Accumulator Line/Pressurizer Surge Line Data -9* -9*
2. MONTECF Decay Heat Uncertainty Error 0* 0*

D. CURRENT PLANNED PLANT CHANGE EVALUATIONS

1. None 0 0 E. TOTAL RESULTANT LBLOCA PCT Total 1948* 1948*

The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points. The Analysis of Record PCT results reflect the Replacement Steam Generators analysis values.

  • See References 1 and 4 E2-10 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 TABLE 2 JOSEPH M. FARLEY NUCLEAR PLANT TOTAL RESULTANT SMALL-BREAK LOCA PCT (fF)

A. SBLOCA ECCS MODEL ANALYSIS-OF-RECORD UNIT 1 UNIT 2

1. ECCS Analysis 1883* 1868**
2. Burst and Blockage / Time in Life 137* 120**

Total Analysis-of-Record 2020* 1988*

B. PRIOR SBLOCA ECCS MODEL ASSESSMENTS I. Prior 10 CFR 50.46 Assessments Reported as Significant -62* 0

2. Prior 10 CFR 50.59 Assessments A. Addition of Permanent Storage Boxes in Containment 0* 0 B. Finalization of Replacement Steam Generator Data -62# 0 C. Annular Pellet Blanket 10* 0 Sum of Prior Assessments -114* 0 C. CURRENT SBLOCA ECCS MODEL ASSESSMENTS
1. NOTRUMP Mixture Level Tracking / Region Depl Errors 13* **
2. Associated change in Burst and Blockage 12* **

D. CURRENT PLANNED PLANT CHANGE EVALUATIONS

1. None 0 0 E. TOTAL RESULTANT SBLOCA PCT Total 1931* 1988**

The PCT values are rounded up to the next highest integer number to avoid reporting in decimal points.

  • See References 1 and 4
    • The revised analysis-of-record reflects the Unit 2's conversion of downflow to upflow configuration (see References 1 and 3).
  1. See Reference 10 E2-11

Enclosure 3 Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003

  • 1 E Enclosure 3 Vogtle Electric Generating Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 BACKGROUND In accordance with 10 CFR 50.46(aX3Xii), this annual report summarizes the nature of and estimated effect of any changes or errors in the emergency core cooling system (ECCS) model for the period from January 1, 2003 through December 31, 2003 for Vogtle Electric Generating Plant Units 1 and 2.

DISCUSSION 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 Annual Report submitted on June 2, 2003 (Reference 10).

This report has been prepared in accordance with the methodology presented in WCAP-13451 (Reference 1) and additional guidance provided by Westinghouse (Reference 2). The LBLOCA and SBLOCA analyses, Evaluation Model assessments, and planned plant change evaluation results reported herein will be included in a future VEGP Final Safety Analysis Report (FSAR) update.

Large-Break LOCA LBLOCA ECCS MODEL ANALYSIS-OF-RECORD In the Annual Report submitted on June 2, 2003 (Reference 10), SNC reported a LBLOCA PCT of 2040.5 OF for both Unit 1 and Unit 2. This value is based on fuel designs containing 128 Integral Fuel Burnable Absorber (IFBA) rods.

The LBLOCA analysis was performed with the 1981 Version of the Westinghouse ECCS Evaluation Model using BASH (Reference 3) including changes in the methodology for execution of the model described in References 4 and 5, and the latest acceptable LOCBART model. The limiting size break analysis assumes the following information important to the LBLOCA analyses:

o 17x17 VANTAGE+ Fuel Assembly o Core Power = 1.02

  • 3565 MWt o Vessel Average Temperature = 570.7 TF o Steam Generator Plugging Level = 10%

o FQ=2.50 o FAH = 1.65 E3-1

6 *i It Enclosure 3 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 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 CD = 0.6. The LBLOCA LOCBART analysis-of-record calculated PCT value is 2040.5 TF for both Unit 1 and Unit 2.

PRIOR LBLOCA ECCS MODEL ASSESSMENTS There are no LBLOCA prior assessments.

CURRENT LBLOCA ECCS MODEL ASSESSMENTS There are no LBLOCA current assessments.

CURRENT PLANNED PLANT CHANGE EVALUATIONS There are no current planned plant changes that affect PCT for Unit 1 and Unit 2.

TOTAL RESULTANT LBLOCA PCT For Unit 1, the absolute sum of the LBLOCA PCT assessments is 0 'F.

For Unit 2, the absolute sum of the LBLOCA PCT assessments is 0 'F.

E3-2 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 UNIT 1 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. LBLOCA ECCS MODEL ANALYSIS-OF-RECORD LOCBART Analysis Result (128 IFBA) 2040.5 OF B. PRIOR LBLOCA ECCS MODEL ASSESSMENTS Combined assessments previously reported as significant +0 OF Combined planned plant change evaluations +0 OF C. CURRENT LBLOCA ECCS MODEL ASSESSMENTS None +0 OF D. CURRENT PLANNED PLANT CHANGE EVALUATIONS None +0 OF E. TOTAL RESULTANT LBLOCA PCT Total 2040.5 F E3-3

  • n -U Enclosure 3 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 UNIT 2 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. LBLOCA ECCS MODEL ANALYSIS-OF-RECORD LOCBART Analysis Result (128 IFBA) 2040.5 OF B. PRIOR LBLOCA ECCS MODEL ASSESSMENTS Combined assessments previously reported as significant +0 OF Combined planned plant change evaluations +0 OF C. CURRENT LBLOCA ECCS MODEL ASSESSMENTS None +0 OF D. CURRENT PLANNED PLANT CHANGE EVALUATIONS None +0OF E. TOTAL RESULTANT LBLOCA PCT Total 2040.5 F E34 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2003 Small-Break LOCA SBLOCA ECCS MODEL ANALYSIS-OF-RECORD In the Annual Report submitted on June 2, 2003 (Reference 10), SNC reported a SBLOCA PCT of 1138.0 TF for both Unit 1 and Unit 2. 1 The SBLOCA analysis was performed with the Westinghouse ECCS Evaluation Model using NOTRUMP (References 6 and 7), including changes to the methodology described in References 8 and 9, and the latest acceptable SBLOCTA model. The limiting size break analysis assumes the following information important to the SBLOCA analyses:

o 17x17 VANTAGE+ Fuel Assembly o Core Power = 1.02

  • 3565 MWt o Vessel Average Temperature = 570.7 "F o Steam Generator Plugging Level = 10%

o FQ =2.58 o FAH = 1.70 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 SBLOCTA analysis-of-record calculated PCT value is 1138.0 "F for both Unit 1 and Unit 2.

PRIOR SBLOCA ECCS MODEL ASSESSMENTS There are no SBLOCA prior assessments.

CURRENT SBLOCA ECCS MODELASSESSMENTS There are no SBLOCA current assessments.

CURRENT PLANNED PLANT CHANGE EVALUATIONS There are no current planned plant changes that affect PCT for Unit 1 and Unit 2.

TOTAL RESULTANT SBLOCA PCT For Unit 1, the absolute sum of the SBLOCA PCT assessments is 0 'F.

For Unit 2, the absolute sum of the SBLOCA PCT assessments is 0 "F.

E3-5 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 UNIT 1 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. SBLOCA ECCS MODEL ANALYSIS-OF-RECORD SBLOCTA Analysis Result 1138.0 OF B. PRIOR SBLOCA ECCS MODEL ASSESSMENTS Combined assessments previously reported as significant +0 O0 F Combined planned plant change evaluations + 0 OF C. CURRENT SBLOCA ECCS MODEL ASSESSMENTS None + 0 OF D. CURRENT PLANNED PLANT CHANGE EVALUATIONS None + 0 OF E. TOTAL RESULTANT SBLOCA PCT Total 1138.0 'F E3-6 10 CFR 50A6 ECCS Evaluation Model Annual Report for 2003 UNIT 2 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. SBLOCA ECCS MODEL ANALYSIS-OF-RECORD SBLOCTA Analysis Result 1138.0 'F B. PRIOR SBLOCA ECCS MODEL ASSESSMENTS Combined assessments previously reported as significant + 0 'F Combined plant change evaluations + 0 OF C. CURRENT SBLOCA ECCS MODEL ASSESSMENTS None + 0 OF D. CURRENT PLANNED PLANT CHANGE EVALUATIONS None + 0 OF E. TOTAL RESULTANT SBLOCA PCT Total 1138.0 'F E3-7

4 094 Enclosure 3 10 CFR 5OA6 ECCS Evaluation Model Annual Report for 2003 CONCLUSION As documented in the preceding tables, the updated VEGP large-break and small-break LOCA analyses PCTs -remain in compliance with 10 CFR 50.46(bX 1), specifically requiring that the PCT shall not exceed 2200 'F. As such, there is no need for reanalysis or taking any other actions in accordance with 10 CFR 50.46(aX3Xii) because compliance with 10 CFR 50.46(bX)1) has been maintained.

REFERENCES

1. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"

October 1992.

2. Westinghouse letter GP-I 7337, "Southern Nuclear Operating Company, Inc., Vogtle Electric Generating Plant Units I and 2, 10 CFR 50.46 Annual Notification Reporting for 2001 ," March 1,2002.
3. "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code,"

WCAP-10266-P-A, Revision 2 (Proprietary) and WCAP-11524-A, Revision 2 (Non-Proprietary), March 1987.

4. Westinghouse letter NTD-NRC-94-4 143 from N. J. Liparulo to W. T. Russell (USNRC),

"Change in Methodology for Execution of BASH Evaluation Model," May 23, 1994.

5. Westinghouse letter NTD-NRC-95-4540 from N. J. Liparulo to W. T. Russell (USNRC),

"Change in Methodology for Execution of BASH Evaluation Model," August 29, 1995.

6. "NOTRUMP: A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Non-Proprietary), August 1985.
7. "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, August 1985.
8. "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," WCAP- 10054-P-A, Addendum 2, Revision 1, July 1997.
9. "Model Changes to the Westinghouse Appendix K Small Break LOCA NOTRUMP Evaluation Model: 1988 - 1997," WCAP-15085, July 1998.
10. NL-03-0999, "10 CFR 50.46 ECCS Evaluation Model Annual Reports for 2002," (multi-docket) letter from J. B. Beasley, Jr. (SNC) to USNRC, June 2, 2003.

E3-8

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