LIC-14-0062, Annual Report for 2013 Loss-of-Coolant Accident (LOCA) / Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46

Annual Report for 2013 Loss-of-Coolant Accident (LOCA) / Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46

ML14118A208
Person / Time
Site:	Fort Calhoun
Issue date:	04/25/2014
From:	Prospero M Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-14-0062
Download: ML14118A208 (6)

Text

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ii;;;;ii Omaha Public Power District 444 South 1ffh Street Mall Omaha, NE 68102-2247 10 CFR 50.46 April 25, 2014 U C-14-0062 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Fort Calhoun Station, Unit No.1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285

Subject:

Annual Report for 2013 Loss-of-Coolant Accident (LOCA) / Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46

References:

1. EMF-21 03(P)(A) , Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," Framatome ANP, Inc., April 2003

2. EMF-2328(P)(A), Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," Framatome ANP, Inc., March 2001

3. Letter from OPPD (L. P Cortopassi) to NRC (Document Control Desk),

"30-Day Report of a Significant Change in the Loss-of-Coolant Accident (LOCA)/Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46," dated September 20,2013 (UC-13-0133)

In accordance with 10 CFR 50.46(a)(3)(ii), the Omaha Public Power District (OPPD) hereby submits the annual 10 CFR 50.46 summary report for 2013. This summary report provides an update of all identified changes or errors in the LOCAlECCS codes, methods, and applications used by AREVA to model Fort Calhoun Station (FCS) , Unit No.1. References 1 and 2 respectively describe the Large Break (LB) and Small Break (SB) LOCA analysis methodology used by AREVA for the FCS Analyses of Record (AOR) .

One LB LOCA Analysis Peak Clad Temperature (PCT) Model Assessment error of +6°F was discovered in 2013. This error is described in Attachment 1. Attachment 2 provides the 2013 LB LOCA Margin Summary Sheet for FCS. As a result of the -55°F total errors reported in previous years, the LB LOCA PCT changed from the baseline value of 1636°F (as reported in the FCS Updated Safety Analysis Report) to 158]oF. The sum of the absolute value of the errors/changes in the LB LOCA AOR is 91°F.

U. S. Nuclear Regulatory Commission LlC-14-0062 Page 2 One SB LOCA Analysis PCT Model change of +309°F was evaluated in 2013. This change was previously reported in Reference 3 and is described in Attachment 3. Attachment 4 provides the 2013 SB LOCA Margin Summary Sheet for FCS. As a result of the -100°F total errors reported in previous years, the SB LOCA PCT changed from the baseline value of 1537°F (as reported in the FCS Updated Safety Analysis Report) to 1746°F. The sum of the absolute values of the errors/changes in the SB LOCA AOR is 417°F.

In summary, the FCS PCT values for SB and LB LOCA continue to remain significantly less than the 10 CFR 50.46(b)(1) acceptance criterion of 2200°F. Current plans are to submit a new RLBLOCA analysis using Revision 3 of Reference 1 within one year of its approval by the NRC, and a new SBLOCA analysis using Supplement 1 of Reference 2 within one year of approval by the NRC, as previously reported (AR60118).

If you should have any questions, please contact Mr. Bill Hansher at (402) 533-6894.

No new commitments to the NRC are made in this letter.

Respectfully, Michael J. Prospero Plant Manager MJP/TAH/brh Attachments: 1. 10 CFR 50.46 Large Break LOCA Model Assessments

2. Large Break LOCA Margin Summary Sheet -Annual Report

3. 10 CFR 50.46 Small Break LOCA Model Assessments

4. Small Break LOCA Margin Summary Sheet -Annual Report c: M. L. Dapas, NRC Regional Administrator, Region IV J. M. Sebrosky, NRC Senior Project Manager J. C. Kirkland, NRC Senior Resident Inspector LlC-14-0062 Page 1 10 CFR 50.46 Large Break LOCA Model Assessments Issue with S-RELAP5 routine associated with the RODEX3a fuel rod model While performing code restructuring activities a code developer reported an issue in an S-RELAP5 routine associated with the RODEX3a fuel rod model in the code.

In realistic large break loss of coolant accident (RLBLOCA) analyses, RODEX3a is used to calculate the fuel rod conditions. The issue involves the trapped stack model in subroutine mdatr3, which is part of the RODEX3a fuel rod model in the code. The error affects any RODEX3a based S-RELAP5 analysis which contains a "trapped stack" of fuel pellets. A "trapped stack" condition exists in any fuel rod containing a "locked" gap with open gaps lying at lower axial levels. A gap is locked when the calculated gap dimension is less than 0.5 mil. That dimension was chosen for the locked criteria to account for roughness, pellet cocking, and cladding ovality effects. All axial levels below the lowest locked gap are part of a trapped stack.

The erroneous coding in mdatr3 involves an incorrect variable addressing which essentially deactivates the trapped stack model. The effect of this error would not be obvious in existing analyses since preliminary assessments indicate the effect of a functioning trapped stack model is very small. Although the effect is small it was determined that it can be conservative or non-conservative depending of the steady-state initial stored energy.

A development version of S-RELAP5 was prepared with the correct evaluation of the trapped stack model and several code validation and plant sample problems were repeated. The assessments included analyses for RLBLOCA Rev O. The SBLOCA analysis is not affected by this change because RODEX2, as opposed to RODEX3a, is used in the analysis.

The estimated impact of this change on the Fort Calhoun RLBLOCA analysis calculated peak cladding temperature is +6°F.

LI C-14-0062 Page 1 Large Break LOCA Margin Summary Sheet -Annual Report Plant Name: Fort Calhoun Station, Unit No.1 Utility Name: Omaha Public Power District Evaluation Model: Large Break LOCA Net PCT Absolute Effect (DPCT) PCT Effect Prior 10 CFR 50.46 Changes or Error A. -55°F 85°F Corrections-Previous Years Prior 10 CFR 50.46 Changes or Error B. Corrections-This Year

+6°F 6°F Absolute Sum of 10 CFR 50.46 Changes 91°F The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of the PCT impact for changes and errors identified since this analysis is less than 2200°F.

LI C-14-0062 Page 1 10 CFR 50.46 Small 8reak LOCA Model Assessments Reduction of High Pressure Safety Injection (HPSI) Delivery Flow Changes for S8 LOCA Analysis The SB LOCA analysis has been re-evaluated to quantify the impact of the HPSI flow rates reduction.

In the Analysis of Record (AOR), the limiting break size reported corresponded to 3.5 in.

diameter. With the HPSI flow reduction, the new limiting break size was identified as 3.0 in.

diameter. The smaller breaks deploy the safety injection tank (SIT) later, due to the slower primary system depressurization; this effect compounded with the reduction in HPSI flow, both contribute to the rise in the limiting PCT.

The PCT impact of the HPSI flow reduction for Fort Calhoun SB LOCA AOR is +30goF and was previously reported in Reference 3.

II C-14-0062 Page 1 Small Break LOCA Margin Summary Sheet -Annual Report Plant Name: Fort Calhoun Station, Unit No.1 Utility Name: Omaha Public Power District Evaluation Model: Small Break LOeA Net PCT Absolute Effect (0 PCT) PCT Effect Prior 10 CFR 50.46 Changes or Error A. -100°F 108°F Corrections-Previous Years Prior 10 CFR 50.46 Changes or Error B. +309°F 309°F Corrections-This Year Absolute Sum of 10 CFR 50.46 Changes 417°F The sum of the PCT from the most recent analysis using an acceptable evaluation model and the estimates of the PCT impact for changes and errors identified since this analysis is less than 2200°F.