NL-12-1983, 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2011 and Significant Change/Error Report
| ML12293A093 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/18/2012 |
| From: | Ajluni M Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-12-1983 | |
| Download: ML12293A093 (10) | |
Text
Mark J. A;luni, P.E. Southern Nuc:lear Nuclear Licensing Director Operating Company, Inc:.
40 Inverness Center Parkway Post Office Box 1295 Birm ingham, Alabama 35201 Tel 205 992.7673 Fax 205.992.7885 SOUTHERN'\'
COMPANY October 18, 2012 Docket Nos.: 50-348 NL-12-1983 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2011 and Significant Change/Error Report Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.46(a)(3)(ii), Southern Nuclear Operating Company (SNC) is submitting the enclosed Joseph M. Farley Nuclear Plant Units 1 & 2 (FNP) emergency core cooling system (ECCS) evaluation model significant change/error report.
On September 20,2012, SNC received notification from its fuel vendor, Westinghouse Electric Company LLC (Westinghouse), of a change/error report which included the estimated effect on the FNP limiting ECCS analysis fuel peak cladding temperature (PCT). Westinghouse performed an evaluation for the FNP best estimate large break loss-of-coolant accident (BE LBLOCA) analysis of record (AOR), and concluded that the estimated PCT impact is 150°F for 10 CFR 50.46 reporting purposes. This constitutes a significant change/error requiring a 30 day report, which is provided in Enclosure 1. Details of this evaluation are provided in Enclosure 2.
The resulting FNP BE LBLOCA PCT continues to meet the criterion of 10 CFR 50.46(b)(1) (i.e., =::; 2200 OF) with sufficient margin such that no reanalysis is required .
The requirements of 10 CFR 50.46(a)(3)(ii) also include an annual report by licensees of changes to, or errors in, acceptable ECCS models that affect the temperature calculation and the estimated effect of changes or errors on the limiting ECCS analysis. Accordingly, Enclosure 1 also includes a summary of changes/errors for 2011 for both the FNP LBLOCA and the FNP small-break loss-of-coolant accident (SBLOCA).
U. S. Nuclear Regulatory Commission NL-12-1983 Page 2 This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.
Respectfully submitted, M. J. Ajluni Nuclear Licensing Director MJAJRMJ/lac
Enclosures:
- 1. Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2011 & Significant Change/Error Report
- 2. Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown cc: Southern Nuclear Operating Company Mr. S . E. Kuczynski, Chairman , President & CEO Mr. D . G . Bost, Executive Vice President & Chief Nuclear Officer Mr. T. A. Lynch, Vice President - Farley Mr. 8. L. Ivey, Vice President - Regulatory Affairs Mr. 8. J. Adams, Vice President - Fleet Operations RTYPE : CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley
Farley Nuclear Plant 10 50.46 ECCS Evaluation Model Annual Report for 2011 & Significant Change/Error Report Enclosure 1 Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2011 & Significant Change/Error Report to NL-12-1983 Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model for 2011 & Significant Change/Error Report Joseph M. Farley Nuclear Plant - Unit 1 ASTRUM Best Estimate Large Break LOCA Evaluation Table 1-1 a - Current Description (6 PCT) pellet thermal conductivity degradation (TCD) and factor burndown were not explicitly considered in Joseph M.
Farley Units 1 and 2 (FNP) best estimate large break coolant accident (BE LBLOCA) analysis of record (AOR). A quantitative evaluation was performed to assess 1 cladding temperature (PCT) effect of fuel pellet TCD and peaking factor burndown on the FNP BE LBLOCA analysis and concluded that the estimated PCT impact is 150°F for 10 CFR 50.46 reporting purposes.
Table 1-1 b - Summary of Changes/Errors (2011)
Description None affecting PCT Table 1-1c Cumulative Impact of Description ed (Reference 1) 18 of Changes/Errors (2011) (Reference 2) o Changes/Errors (Reference 3) 1 Cumulative E1-1 to NL~1 983 Farley Nuclear Plant 10 50.46 Model Annual Report 2011 & Significant Change/Error Report Joseph M. Farley Nuclear Plant - Unit 1 NOTRUMP Appendix K Small Break lOCA Evaluation Table 1-1 d - Summary of Changes/Errors (2011) I Estimated Description Effect
(.1 PCT)
None affecting PCT N/A Table 1 e Cumulative Impact of Changes/Errors
.1IPCTI I co F)
.. Changes/Errors Previously Reported (Reference 1) 0
.. Summary of Changes/Errors (2011) (Reference 2) 0 Cumulative Total 0
-2
983 Farley Nuclear 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2011 & Significant Change/Error Report Joseph M. Farley Nuclear Plant - Unit 2 ASTRUM Estimate large lOCA Evaluation Table 1-2a - Current Changes/Errors Estimated Description (Ll PCT) pellet thermal conductivity degradation (TCD) and peaking factor burndown were not explicitly considered in the M.
Farley Units 1 and 2 (FNP) estimate large break loss-of coolant accident LBLOCA) analysis of record (AOR). A quantitative evaluation was performed to assess peak 150 cladding temperature (PCT) effect of fuel pellet and peaking factor burndown on the FI\IP BE LBLOCA analysis and concluded that impact is 150°F 10 CFR reporting Table 1 Summary of Changes/Errors (2011)
Estimated E!LII:' ...
Description R;;;.ln:::",
(Ll PCT)
None affecting PCT N/A Table 1-2c - Cumulative Impact of Changes/Errors PCTI Description (OF)
I
- Changes/Errors Previously Reported (Reference 1) 18
- Summary of Changes/Errors (2011) (Refefence 2) 0
- Current Changes/Errors (Reference 3) 1 Cumulative Total 168
-3 NL-12-1983 Farley Nuclear Plant 10 CFR Evaluation Model Annual for 2011 &
Joseph M. Farley Nuclear Plant - Unit 2 NOTRUMP Appendix K Small Break LOCA Evaluation Table 1-2d Summary of Changes/Errors (2011)
Estimated Description Effect
(~ PCn None affecting PCT N/A Table 1-2e - Cumulative Impact of Changes/Errors Description ~IPC"I (OF)
- Previously Ri;;;tJUf L!;;U (Reference 1) 0
- Summary of Changes/Errors (2011) (Reference 0 Cumulative Total 0
-4 NL-1 983 Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual for 2011 & Significant Change/Error Report
References:
- 1. Letter from Mark J. Ajluni to USNRC (NL-11-2402), "Edwin I. Hatch Nuclear Plant, Joseph M. Farley Nuclear Plant, Vogtle Electric Generating Plant 10 50.46 Evaluation Model Annual Reports for 2010," dated December , 2011 Westinghouse Letter LTR-LlS-12-11 "J. M. Farley Units 1 and 2,10 CFR 50.46 Annual Notification and Reporting for 2011," dated 23, 2012
- 3. Westinghouse Notification Letter LTR-US-12-407, "J. M. Farley Units 1 and 2 10 50.46 Notification and Reporting for Pellet Thermal Conductivity Degradation and Peaking Burndown," dated September 20,2012
Farley Nuclear Plant 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2011 & Significant Change/Error Report Enclosure 2 Evaluation of Fuel Pellet Thermal Conductivity Degradation And Peaking Factor Burndown to NL-12-1983 Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown
Background
Fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown were not explicitly considered in the Joseph M. Farley Nuclear Plant Units 1 & 2 (FNP) best estimate large break loss-of-coolant accident (BE LBLOCA) analysis of record (AOR). Nuclear Regulatory Commission (NRC) Information Notice 2011-21 notified addressees of recent information obtained concerning the impact of irradiation on fuel thermal conductivity and its potential to cause significantly higher peak cladding temperature (PCT) results in realistic emergency core cooling system (ECCS) evaluation models. This evaluation provides an estimated effect of fuel pellet TCD and peaking factor burndown on the PCT calculation for the FNP BE LBLOCA AOR.
Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)
Estimated Effect A quantitative evaluation was performed to assess the PCT effect of fuel pellet TCD and peaking factor burndown on the FNP BE LBLOCA analysis and concluded that the estimated PCT impact is 150°F for 10 CFR 50.46 reporting purposes. The peaking factor burndown included in the evaluation is provided in Table 1 and is conservative for the current cycle. SNC and Westinghouse utilize processes which ensure that the LOCA analysis input values conservatively bound the as-operated plant values for those parameters and will be validated as part of the reload design process.
Table 1: Peaking Factors Assumed in the Evaluation of TCO Rod Burnup FdH (1,2) Fa Transient(1) Fa Steady-State (GWO/MTU) 0 1.700 2.500 2.050 30 1.700 2.500 2.050 60 1.400 1.829 1.500 62 1.400 1.829 1.500
- 1) Includes uncertainties
- 2) Hot assembly average power follows the same burndown since it is a function of FdH FdH = Nuclear Enthalpy Rise Hot Channel Factor Fa = Heat Flux Hot Channel Factor E2-1