ML12121A006

Annual Report for 2011 Loss-of-Coolant Accident (Loca)/Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46

ML12121A006
Person / Time
Site:	Fort Calhoun
Issue date:	04/27/2012
From:	Herman J Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LlC-12-0053
Download: ML12121A006 (7)

Text

Omaha Public Power Oistrict 444 South 16th Street Mall Omaha, NE 68102-2247 April 27, 2012 LlC-12-0053 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

References:

1. Docket 50-285.

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

March 2001.

3. EMF-2103(P)(A), Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," Framatome ANP, Inc., April 2003.

4. Letter from OPPD (R.P. Clemens) to NRC (Document Controll Desk), 30 Day Report of a Significant Change/Error in the Loss of Coolant Accident (LOCA)/Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46, dated April 8, 2009 (LlC-09-0028) .

Subject:

Annual Report for 2011 Loss-of-Coolant Accident (LOCA)/Emergency Core Cooling System (ECCS) Models Pursuant to 10 CFR 50.46 The Omaha Public Power District (OPPD) has received the 2011 AREVA (formerly Framatome ANP) 10 CFR 50.46 Annual Notification Report for the Small Break (SB) and Large Break (LB) LOCA Analyses that are subject to the reporting requirements of 10 CFR 50.46. Therefore, in accordance with 10 CFR 50.46(a)(3)(ii), OPPD submits the Annual 10 CFR 50.46 Summary Report. This report updates all identified changes or errors in the LOCA/ECCS codes, methods, and applications used by AREVA to model Fort Ca,lhoun Station (FCS),

Unit No.1. References 2 and 3 respectively describe the SB and LB LOCA analysis methodology used by AREVA for the FCS Analyses of Record (AOR).

Attachment 1 discusses one (1) 10 CFR 50.46 Model Assessment error with a Peak Clad Temperature (PCT) impact of -32°F that was discovered in the SB LOCA Analysis in 2011. Attachment 2 provides the 2011 SB LOCA Margin Employment with Eq ual Opportun ity

- - - --------------~----,

U. S. Nuclear Regulatory Commission LlC-12-0053 Page 2 Summary Sheet for FCS. In combination with the -68°F total errors from previous years, this error changes the SB LOCA PCT from the baseline val'ue (153rF) reported in the FCS Updated Safety Analysis Report (USAR) to 143rF.

The sum of the absolute values of the errors/changes in the SB LOCA AOR is 108°F. discusses two (2) 10 CFR 50.46 Model Assessment errors with a PCT impact of +8°F (O°F and +8°F respectively) that were discovered in the LB LOCA Analyses in 2011. Attachment 4 provides the 2011 LB LOCA Margin Summary Sheet for FCS. In combination with the -62°F total errors from previous years, these errors change the LB LOCA PCT from the baseline value (1636°F) reported in the FCS USAR to 1582°F. The sum of the absolute value of the errors/changes in the LB LOCA AOR is 84°F.

In summary, the FCS PCT values for the SB and LB LOCA continue to remain significantly less than the 10 CFR 50.46(b)(1) acceptance criterion of 2200°F.

OPPD has previously (Reference 4) communicated its intent to update the SB and LB LOCA Analyses for a planned Extended Power Uprate (EPU) of FCS.

However, the EPU project has been delayed indefinitely and since a significant margin to the 2200°F limit exists, OPPD has decided not to reanalyze the SB and LB LOCA events.

No commitments to the NRC are made in this letter. If you should have any questions, please contact Mr. Bill Hansher at (402) 533-6894.

Sincerely, J . B. Herman Division Manager-Nuclear Engineering Attachments:

1. 10 CFR 50.46 Small Break LOCA Model Assessments

2. Small Break LOCA Margin Summary Sheet - Annual Report

3. 10 CFR 50.46 Large Break LOCA Model Assessments

4. Large Break LOCA Margin Summary Sheet - Annual Report c: E. E. Collins, Jr., NRC Regional Administrator, Reg'ion IV L. E. Wilkins, NRC Project Manager J. C. Kirkland, NRC Senior Resident Inspector LlC-12-0053 Page l' 10 CFR 50.46 Small Break LOCA Model Assessments Sieicher-Rouse Heat Transfer Correlation in S-RELAP5 Sieicher-Rouse is one of the correlations used to define the heat transfer between the fue.1and coolant. This correlation is applicable to both Large Break (LB) and Small Break (SB) analyses performed with the S-RELAP5 computer code .

During development by AREVA of a boiling water reactor (BWR) loss-of-coolant accident (LOCA) methodology based on S-RELAP5, the behavior of the Sieicher Rouse correlation relative to other single-phase vapor heat transfer correlations was reviewed and it was questioned whether the Sieicher-Rouse correlation was correct. It was also discovered that another industry code uses the Sieicher p Rouse correlation, but the form for the correlation is different from that used in the S-RELAP5 implementation of Sieicher-Rouse. The concern is related to the form of the equation for calculating the exponent of the temperature ratio correction term.

The S-RELAP5 form is: n =-IOglO (TwfTg) 1/4 + 0.3 The alternate form used in another industry code is: n = -[IOglO (TwfTg)l1 /4 + 0.3 The alternate form is more consistent with other heat transfer correlations and expected physical trends.

Assessments of the potential impact of using the alternate Sieicher-Rouse correlation form were performed. A development version of S-RELAP5 was prepared with the alternate Sieicher-Rouse form and several code validation and plant sample problems were repeated. Additional analyses were performed using a different heat transfer correlation for single-phase vapor heat transfer.

The assessments included analyses for both Realistic LB LOCA Revision 0 and pressurized water reactor (PWR) S-RELAP5 SB LOCA methodologies.

The Peak Clad Temperature (PCT) impact for the Fort Calhoun Station SB LOCA is -32°F.

Attachment 2 LlC-12-0053 Page 1 Small Break LOCA Margin Summary Sheet - Annual Report Plant Name: Fort Calhoun Station Utility Name: Omaha Public Power District Evaluation Model: Small Break LOCA NetPCT Absolute I I Effect (6PCT) PCT Effect A. Prior 10 CFR 50.46 Changes or Error -68°F 76°F Corrections-Previous Years I B. I Prior 10 CFR 50.46 Changes or Error I -32 °F 32°F Corrections-This Year Absolute Sum of 10 CFR 50.46 108°F C. II II Changes 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.

LlC-12-0053 Page 1 10 CFR 50.46 Large Break LOCA Model Assessments Sieicher-Rouse Heat Transfer Correlation in S-RELAP5 Sieicher-Rouse is one of the correlations used to define the heat transfer between the fuel and coolant. This correlation is applicable to both Large Break (LB) and Small Break (SB) analyses performed with the S-RELAP5 computer code.

During development by AREVA of a boiling water reactor (BWR) loss-of-coolant accident (LOCA) methodology based on S-RELAP5, the behavior of the Sieicher Rouse correlation relative to other singl'e-phase vapor heat transfer correlations was reviewed and it was questioned whether the Sieicher-Rouse correlation was correct. It was also discovered that another industry code uses the Sieicher Rouse correlation, but the form for the correlation is different from that used in the S-RELAP5 implementation of Sieicher-Rouse. The concern is related to the form of the equation for calculating the exponent of the temperature ratio correction term.

The S-RELAP5 form is: n = -IOglO (Tw/Tg) 1/4 + 0.3 The alternate form used in another industry code is: n = -[IOglO (Tw/Tg)]1/4 + 0.3 The alternate form is more consistent with other heat transfer correlations and expected physical trends.

Assessments of the potential impact of using the alternate Sieicher-Rouse correlation form were performed. A development version of S-RELAP5 was prepared with the alternate Sieicher-Rouse form and several code validation and plant sample problems were repeated. Additional analyses were performed using a different heat transfer correlation for single-phase vapor heat transfer.

The assessments included analyses for both Realistic LB LOCA Revision 0 and pressurized water reactor (PWR) S-RELAP5 SB LOCA methodologies.

The Peak Clad Temperature (PCT) impact for the Fort Calhoun Station LB LOCA is +8°F.

LB LOCA Model producing Non-physical Phenomena in Upper Plenum The impact of liquid and vapor flow spikes from the upper pjenum (UP) into the hot channel (HC) and surrounding six assembly regions of the core and a nonphysical flow pattern in the upper plenum was evaluated . Even though Counter Current Flow Limit (CCFL) modeling was applied at the HC exit junction, LlC-12-0053 Page 2 it will not be activated due to the negative spikes in steam velocities (from upper plenum to HC).

The current LB LOCA reactor vessel modeling was traced back to the EMF-21 03 sample problem for a 3-loop Westinghouse (W) plant. This W 3-loop plant has a geometry feature in the upper plenum known as "flow mixers or standpipes."

Due to this geometry feature, the upper plenum was broken into two sections, one to an open hole region and one to a flow mixer region. The modeling in the sample problem blocked the cross flow between radial junctions in the first level of upper plenum and this was carried through in plants without flow mixers as a methodology conservatism .

The UP nodalization for these plant cases was revised to make it consistent with the geometry. In addition, in all plant cases, a high reverse loss coefficient is applied to the HC and central core to UP junctions at the beginning of the core reflooding phase. Cases were rerun that had liquid down flow and potentially affect the Analyses of Record (AOR) PCT limit.

The PCT impact for the Fort Calhoun Station LB LOCA is O°F.

Attachment 4 LlC-12-0053 Page 1 Large Break LOCA Margin Summary Sheet - Annual Report Plant Name: Fort Calhoun Station Utility Name: Omaha Public Power District Evaluation Model: Large Break LOeA Net PCT Effect Absolute

(~PCT) PCT Effect I A. Prior 10 CFR 50.46 Changes or Error -62°F 76°F

, Corrections-Previous Years B. I Prior 10 CFR 50.46 Changes or Error +8°F 8°F II Corrections-This Year C. Absolute Sum of 10CFR 50.46 Changes I 84°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.