JAFP-04-0135, James a FitzPatrick, 10 CFR 50.46 Annual Report - Errors in Emergency Core Cooling System (ECCS) Evaluation Models

James a FitzPatrick, 10 CFR 50.46 Annual Report - Errors in Emergency Core Cooling System (ECCS) Evaluation Models

ML042400357
Person / Time
Site:	FitzPatrick
Issue date:	08/23/2004
From:	Ted Sullivan Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-04-0135
Download: ML042400357 (8)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. Fitzpatrick NPP Entergy RO. Box 110 Lycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 T.A. Sullivan August 23, 2004 Site Vice President - JAF JAFP-04-0135 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

James A FitzPatrick Nuclear Power Plant Docket No. 50-333 10 CFR 50.46 Annual Report - Errors in Emergency Core Cooling System (ECCS) Evaluation Models

REFERENCE:

1. Entergy Nuclear Operations, Inc. letter T.A. Sullivan to USNRC (JAFP-03-0124) dated August 28, 2003 regarding same subject.

Dear Sir:

The attached report summarizes changes and errors in emergency core cooling system (ECCS) evaluation models in accordance wvith 10 CFR 50.46(a)(3)(ii) for the period July 1, 2003 to June 30, 2004 for Entergy's James A. FitzPatrick Nuclear Power Plant. Three changes or errors that should have been reported in Reference 1, but were inadvertently left out of that report are also included.

A total of two changes and two errors to the FitzPatrick model have been identified since the last report (Reference 1). Thirty day reports were not submitted because none of the changes or errors qualify as a significant change (a peak clad temperature change of greater than 500 F) according to 10 CFR 50.46(a)(3)(i).

Corrected for these changes and errors, estimated peak clad temperatures (PCTs) are unchanged from the previous reports and remain below the 22000 F requirement of 10 CFR 50.46(b)(1).

This letter contains no new commitments. If you have any questions, please contact Mr. Richard Plasse at (315) 349-6793.

Very truly yours, WAe. - - 4 - - D C T. A. SULLIVAN No I

Attachment:

10 CFR 50.46(a)(3)(ii) Annual Report on Changes and Errors in Emergency Core Cooling System (ECCS) Evaluation Models for the period from July 1,2003 to June 30, 2004.

cc: Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Office of the Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Patrick Milano, Project Manager Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop: 8-C2 Washington, DC 20555

Attachment to JAFP-04-0135 INTRODUCTION This report summarizes changes and errors in emergency core cooling system (ECCS) evaluation models in accordance with 10 CFR 50.46(a)(3)(ii) for the period July 1, 2003 to June 30, 2004 for Entergy's James A. FitzPatrick Nuclear Power Plant.

A total of two (2) changes and two (2) errors to the FitzPatrick model have been identified since the last annual report (Reference 26). The changes and errors (References 22, 23, 24 and 25) did not result in a peak clad temperature (PCT) change.

Thirty-day reports were not submitted for the two errors because they do not qualify as significant changes (a PCT change of greater than 500 F) according to 10 CFR 50.46(a)(3)(i).

Table I summarizes the changes and errors to the current FitzPatrick ECCS evaluation models. The last four entries represent ECCS evaluation changes or errors identified during the period July 1, 2003 to June 30, 2004 (including three previous inadvertent omissions).

Additional information on the changes and errors identified during the reporting period is provided below.

DISCUSSION OF CHANGES OR ERRORS IDENTIFIED DURING REPORTING PERIOD GESTR Input to SAFER Error (2002-03)

Fuel rod gap conductance is calculated by GESTR as functions of LHGR and EXP and used as input to SAFER. An error in the interpolation code resulted in initial gap conductance slightly lower than it should have been.

This error did not result in a change in PCT. Therefore, this error did not qualify as a significant change according to 10 CFR 50.46(a)(3)(i), and no 30-day report was submitted.

Miuration of SAFER04 from VAX to Alpha Computer Platform (2002-04)

The LOCA evaluation code SAFER04 has been migrated from the VAX computer platform to the Alpha computer platform. The change in computer platform may result in a change in the calculated PCT due to changes in the processor word size and FORTRAN compiler characteristics.

This change in the model did not result in a change in PCT.

WEVOL Code Error(2002-05)

The WEVOL code is used to calculate the weight and volume inputs for SAFER analysis ofjet pump plants. An error was found in the WEVOL code which affects the calculated vessel volume in the downcomer region. The free volume in the region of the shroud head was calculated incorrectly. The code did not properly account for the volume of the Page 1 of 6

Attachment to JAFP-04-0135 standpipes inside the shroud head thickness. This resulted in the value of the downcomer free volume being too small by 4 - 10 ft3 .

This error did not result in a change in PCT. Therefore, this error did not qualify as a significant change according to 10 CFR 50.46(a)(3)(i), and no 30-day report was submitted.

Impact of Postulated HVdrogen-Oxygen Recombination (2003-05)

A new LOCA heat source has been postulated. This heat source involves the recombination of hydrogen and oxygen within the fuel bundle during the core heatup.

The additional heat will raise the temperature of the steam heat sink in the bundle, resulting in a potential increase in the peak cladding temperature and local oxidation.

This recombination is spontaneous at temperatures above approximately 900'F. The hydrogen is generated by the steam-zirconium reaction during heatup. The oxygen enters the vessel either as a dissolved gas in the ECCS water or through the break when the vessel fully depressurizes and draws the containment non-condensable gases back into the vessel. The current LOCA evaluation models do not include this new heat source.

Pending disposition of this phenomenon, a change notification is supplied to provide the impact of hydrogen-oxygen recombination on the cladding temperature and local oxidation.

This change in the model did not result in a change in PCT.

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Attachment to JAFP-04-0135 TABLE 1 - ACCOUNTING OF LICENSING BASIS PEAK CLAD TEMPERATURES FOR FITZPATRICK Report Estimated Updated Period PCT Change PCT ECCS Evaluation GE11 GE12 GE14 GE11 GE12 GE14 Fuel") Fuel Fuel(2) Fuel Fuel Fuel Baseline 1993 FitzPatrick LOCA Baseline N/A N/A 1570°F N/A N/A Evals. Analysis (Ref. 7) I FitzPatrick Reload 12 N/A Base- N/A N/A 1370°F N/A Supplemental Report line (Ref. 8)

FitzPatrick Reload 13 N/A 1570°F 1370°F N/A Supplemental Report Baseline (Ref. 9)

Prior to 10 CFR 50.46 +50°F N/A 1620°F 1420°F N/A July 1, Notification Regarding 2000 Sensitivity To Small Input Parameter Changes (Ref. 10) 10 CFR 50.46 +50 F N/A 1625°F 1425°F N/A Notification Regarding Minor Code Corrections (Ref. 10) 10 CFR 50.46 +100 F N/A 1635°F 1435°F N/A Notification Regarding Bottom Head Drain (Ref. 11)

July 1, 10 CFR 50.46 -5F N/A 1630°F 1430°F N/A 2000 - Notification Regarding June 30, Time Step Size (Ref. 3) 2001 Estimated Effect Of +90°F N/A 1720°F 1520°F N/A Condensation Error On PCT (Ref. 12)

Estimated Effect Of +10°F N/A 1730°F 1530F N/A Pressure Rate Inconsistency Error On PCT (Ref. 13)

Estimated Effect Of N/A -50 F N/A 1730°F 1525°F N/A Accounting Error On PCT (Ref. 14)

July 1, SAFER Core Spray +50 F N/A 1735°F 1530°F N/A 2001 - Injection Elevation Error June 30, (Ref. 16) 2002 Impact of SAFER Bulk +10 0 F N/A 1745°F 1540°F N/A Water Level Error on PCT (Ref. 17)

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Attachment to JAFP-04-0135 TABLE I -ACCOUNTING OF LICENSING BASIS PEAK CLAD TEMPERATURES FOR FITZPATRICK (Continued)

Report Estimated Updated Period PCT Change PCT ECCS Evaluation GE11 GE12 GE14 GE11 GE12 GE14 Fuel Fuel Fuel Fuel Fuel Fuel July 1, Cycle 16 Core Load N/A N/A 1700OF N/A N/A 1700OF 2002 - Using GE 14 Type Fuel June 30, (Ref. 19) 2003 Impact of SAFER N/A 0F OOF N/A 1540WF 1700OF Level/Volume Table Error on the PCT (Ref. 20)

Impact of SAFER Initial N/A 0F N/A N/A 1540OF 1700OF Separator Pressure Drop Error on the PCT (Ref.21)

July 1, Impact of GESTR Input N/A 0F N/A N/A 1540 0 F 1700OF 2003 - File Interpolation Error June 30, on the PCT (Ref. 22)(_)

2004 Impact of SAFER04 N/A 0F N/A N/A 1540 0 F 1700OF Computer Platform Change on the PCT (Ref. 23) (3)

Impact of WEVOL S1 N/A 07F N/A N/A 1540 0 F 1700OF Volume Error on the PCT (Ref. 24) (3)

Impact of Postulated N/A 0°F 0°F N/A 1540°F 1700OF Hydrogen-Oxygen Recombination (Ref. 25)

(1) GEl I fuel was removed from the core at conclusion of Cycle 15, 10/2002 (2) GE14 fuel was first installed during RFI5, 10/2002 (3) This item was inadvertently omitted from the 2002-2003 report Page 4 of 6

Attachment to JAFP-04-0135 REFERENCES

1. NYPA letter, M. J. Colomb to USNRC (JAFP-00-0164), dated July 21, 2000, regarding reporting of changes and errors in ECCS evaluation models.

2. Entergy Nuclear Operations, Inc. letter, M. Kanslerto USNRC, (JPN-01-010) dated June 4, 2001, regarding 10 CFR 50.46(a)(3)(ii) 30-day report, two errors in ECCS evaluation models.

3. Global Nuclear Fuel letter, A. Alzaben (GNF) to C. Franklin (NYPA), (AFA-00-N061 dated November 20, 2000) regarding 10 CFR 50.46 error report - impact of SAFER time step size on the peak clad temperature (PCT) for jet pump plant analyses.

Includes General Electric/Global Nuclear Fuel 10 CFR 50.46 Notification Letter 2000-04 dated November 8, 2000 regarding "Impact of SAFER Time Step Size on the Peak Clad Temperature (PCT) for Jet Pump Plant Analyses." (Proprietary)

4. NEDC-23785-1-PA Rev. 1, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-Of-Coolant Accident Volume II, SAFER - Long Term Inventory Model for BWR Loss-Of-Coolant Analysis," October 1984.

5. NEDC-23785-1-PA Rev. 1, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-Of-Coolant Accident Volume l1l, SAFER/GESTR Application Methodology," October 1984.

6. General Electric Nuclear Energy Report, J11-03757SRL, August 2000, "Supplemental Reload Licensing Report for James A. FitzPatrick, Reload 14, Cycle 15."

7. General Electric Nuclear Energy, "James A. FitzPatrick Nuclear Power Plant SAFER/GESTER-LOCA, Loss-of-Coolant Analysis," Licensing Topical Report NEDC-31317P, Class IlIl (proprietary), Revision 2, April 1993.

8. General Electric Nuclear Energy Report, JI1-02914SRL, Revision 0, August 1996, "Supplemental Reload Licensing Report for James A. FitzPatrick, Reload 12, Cycle 13."

9. General Electric Nuclear Energy Report, J1 1-03359SRL, Revision 1, Class I, October 1998, "Supplemental Reload Licensing Report for James A. FitzPatrick, Reload 13, Cycle 14."

10. General Electric Nuclear Energy, MFN-090-93, June 30, 1993, "Reporting of Changes and Errors in ECCS Evaluation Models."

11. General Electric Nuclear Energy, MFN-020-96, February 20,1996, "Reporting of Changes and Errors in ECCS Evaluation Models."

12. General Electric/Global Nuclear Fuel 10 CFR 50.46 Notification Letter 2001-01 dated May 8, 2001 (via E-mail) regarding "Impact of SAFER Condensation Error on the Peak Clad Temperature (PCT)." (Proprietary)

13. General Electric/Global Nuclear Fuel 10 CFR 50.46 Notification Letter 2001-02 dated May 10, 2001 (via E-mail) regarding "Impact of SAFER Pressure Rate I inconsistency Error on the Peak Clad Temperature (PCT)." (Proprietary)

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Attachment to JAFP-04-0135

14. Global Nuclear Fuel letter, A. Alzaben (GNF) to J. Head (Entergy Nuclear Operations, Inc.), (AFA-01-E004, June 29, 2001) regarding GE12 Upper Bound PCT for FitzPatrick, (Proprietary).

15. Entergy letter, M. Kansler to USNRC, dated August 15, 2001 (JPN-01-0014) regarding "10 CFR 50.46 Annual Report - Errors in Emergency Core Cooling System (ECCS) Evaluation Models."

16. General Electric/Global Nuclear Fuel 10 CFR 50.46 Notification Letter 2002-01 regarding "SAFER Core Spray Injection Elevation Error." (Proprietary)

17. General Electric/Global Nuclear Fuel 10 CFR 50.46 Notification Letter 2002-02 regarding "Impact of SAFER Bulk Water Level Error on the Peak Clad Temperature (PCT)." (Proprietary)

18. Entergy letter, M. Kansler to USNRC, dated August 14, 2002 (JPN-02-023) regarding "10 CFR 50.46 Annual Report - Errors in Emergency Core Cooling System (ECCS) Evaluation Models."

19. Entergy letter, T.A. Sullivan to USNRC, dated November 12, 2002 (JAFP-02-0210) regarding "10 CFR 50.46 Reporting of Changes In Emergency Core Cooling System (ECCS) Evaluation Model."

20. General Electric 10 CFR 50.46 Notification Letter 2003-01, dated May 6, 2003 regarding "Impact of SAFER LevelNolume Table Error on the Peak Cladding Temperature (PCT)." (Proprietary)

21. General Electric 10 CFR 50.46 Notification Letter 2003-03, dated May 6, 2003 regarding "Impact of SAFER Initial Separator Pressure Drop Error on the Peak Cladding Temperature (PCT)." (Proprietary)

22. General Electric 10 CFR 50.46 Notification Letter 2002-03, dated August 26, 2002 regarding "Impact of GESTR Input File Interpolation Error on the Peak Clad Temperature (PCT)." (Proprietary)

23. General Electric 10 CFR 50.46 Notification Letter 2002-04, dated August 26, 2002 regarding uImpact of SAFER04 Computer Platform Change on the Peak Clad Temperature (PCT)." (Proprietary)

24. General Electric 10 CFR 50.46 Notification Letter 2002-05, dated August 26, 2002 regarding 'Impact of WEVOL S1 Volume Error on the Peak Clad Temperature (PCT)." (Proprietary)

25. General Electric 10 CFR 50.46 Notification Letter 2003-05, dated May 13, 2004 regarding "Impact of Postulated Hydrogen-Oxygen Recombination." (Proprietary)

26. Entergy letter, T.A. Sullivan to USNRC, dated August 28, 2003 (JAFP-03-0124) regarding "10 CFR 50.46 Annual Report - Errors in Emergency Core Cooling System (ECCS) Evaluation Models."

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