PLA-7947, 10 CFR 50.46 Annual Report (PLA-7947)

10 CFR 50.46 Annual Report (PLA-7947)

ML21161A005
Person / Time
Site:	Susquehanna
Issue date:	06/10/2021
From:	Cimorelli K Susquehanna, Talen Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7947
Download: ML21161A005 (10)

Text

TALEN~

Kevin Cimorelli Susquehanna Nuclear, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3795 Fax 570.542.1504 ENERGY June 10, 2021 Kevin.Cimorelli@TalenEnergy.com U. S. Nuclear Regulatory Commission 10 CFR 50.46 Document Control Desk Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION 10 CFR 50.46 ANNUAL REPORT Docket Nos. 50-387 PLA-7947 and 50-388 References 1. "Susquehanna Steam Electric Station 10 CFR 50.46-Annual Report (PLA-7870),"

dated June 10, 2020 (ADAMS Accession No. ML20162A0J 4)

2. Framatome Record FSJ-0052508, Revision 1.0, "JO CFR 50.46 PCT Error Reporting/or the Susquehanna Units," dated September 29, 2020

3. Framatome Record FSJ-0054287, Revision 1.0, "JO CFR 50.46 PCT Error Reporting/or the Susquehanna Units," dated Janua,y 19, 2021

4. Framatome Record FSJ-0055556, Revision 1.0, "JO CFR 50.46 PCT Error Reporting/or the Susquehanna Units, " dated April 9, 2021

5. "Susquehanna Steam Electric Station, Units 1 and 2 Issuance ofAmendment Nos. 278 and 260 to Allow Application ofAdvanced Framatome Atrium 11 Fuel Methodologies (EPID L-2019-LLA-0153), "dated Janua,y 21, 2021 (ADAMS Accession No. ML20168B004)

6. ANP-3784P Revision 0, "Susquehanna Units 1 and 2 LOCAAnalysisforATRJUM 11 Fuel," Framatome, June 2019 Pursuant to the reporting requirements of 10 CFR 50.46(a)(3)(ii), Susquehanna Nuclear, LLC is submitting the Emergency Core Cooling System (ECCS) evaluation model annual report for Susquehanna Steam Electric Station (SSES) Units 1 and 2. The attached reports summarize the nature of and estimated effect of any modeling changes or error corrections in the ECCS models. provides a summary of the Framatome EXEM BWR-2000 LOCA (Loss of Coolant Accident) Methodology which is applicable to ATRIUM 10 fuel for the period Febrnary 27, 2020, through April 9, 2021, for SSES Units 1 and 2. Since the last 10 CFR 50.46 annual report dated June 10, 2020 (Reference 1), there were no Peak Cladding Temperature (PCT) changes reported to SSES resulting from a modeling change or error correction to the EXEM BWR-2000 LOCA Methodology. The current licensing basis PCT remains in compliance with 10 CFR 50.46 requirements. provides a summary of the Framatome AURORA-B LOCA Methodology which is applicable to ATRIUM 11 fuel. The AURORA-B LOCA Methodology was approved for use at SSES, Units 1 and 2, in Reference 5. The changes reported herein cover the time period from approval in Reference 5 (i.e., January 21, 2021) through April 9, 2021. Additionally, further changes to the model are described herein which

Document Control Desk PLA-7947 were identified from June 30, 2019, (Reference 6), through January 21, 2021, while AURORA-B was being evaluated by the NRC for use at SSES, Units 1 and 2. Since the last 10 CFR 50.46 annual report dated June 10, 2020, (Reference 1), there were seven non-impacting and three impacting PCT changes reported to SSES resulting from a modeling change or en-or con-ection to the AURORA-B LOCA Methodology. The cun-ent licensing basis PCT remains in compliance with 10 CFR 50.46 requirements.

There are no new or revised regulatory commitments contained in this submittal.

If you have any questions regarding this letter, please contact Ms. Melisa Krick, Manager -

Nuclear Regulatory Affairs, at (570) 542-1818.

Attachments:

1. 10 CFR 50.46 ECCS Evaluation Model Annual Report for ATRIUM 10 Fuel

2. 10 CFR 50.46 ECCS Evaluation Model Annual Report for ATRIUM 11 Fuel Copy: NRC Region I Ms. A. Klett, NRC Project Manager Mr. C. Highley, NRC Senior Resident Inspector Mr. M. Shields, PA DEP/BRP

Attachment 1 to PLA-7947 10 CFR 50.46 ECCS Evaluation Model Annual Report for ATRIUM 10 Fuel

Attachment 1 to PLA-7947 Page 1 of 2 BACKGROUND In accordance with 10 CFR 50.46(a)(3)(ii), this annual report summarizes the nature of and estimated effect of any modeling changes or effor cmTections in the ECCS model for the period February 27, 2020, through April 9, 2021.

DISCUSSION ATRIUM 10 and ATRIUM 11 fuel are currently co-loaded in the SSES Unit 2 Reactor.

ATRIUM 10 is currently loaded in the SSES Unit 1 Reactor; ATRIUM 10 and ATRIUM 11 are expected to be co-loaded in the SSES Unit 1 Reactor following the refueling outage in spring 2022. The ECCS performance evaluation method applicable to ATRIUM 10 fuel for both SSES Units 1 and 2 is the Framatome EXEM BWR-2000 LOCA (Loss of Coolant Accident) Methodology.

For the reporting period of February 27, 2020, through April 9, 2021, there have been no repmiable 10 CFR 50.46 modeling changes or error cmTections to the ECCS evaluation method since the previous 10 CFR 50.46 report (Reference 1).

The total change listed in the last column of Table 1 does not meet the significance threshold for change (50°F) identified in 10 CFR 50.46(a)(3)(i) for which a 30-day repmi is required.

Attachment 1 to PLA-7947 Page 2 of 2 IMPACT Table 1 Non-Zero PCT Changes Resulting from Modeling Changes/ Error Corrections in Calculated ECCS Performance Evaluation Model: Framatome EXEM BWR-2000 LOCA Methodology Estimated Absolute APCT (°F) Value of Description of Change/Error APCT (°F)

HUXY capability enhancement to model each fuel rod individually

-1 1 (ADAMS Accession No. MLI 7158B382)

Updated steam dryer information

+5 5 (ADAMS Accession No. ML19161A131)

Total Since Initial PCT

+4 6 (Reference 4)

CONCLUSION As documented in Table 1, the SSES Units 1 and 2 Loss of Coolant Accident Analysis PCT remains in compliance with 10 CFR 50.46(b)(l), which requires that the PCT not exceed 2200°F.

Attachment 2 to PLA-7947 10 CFR 50.46 ECCS Evaluation Model Annual Report for ATRIUM 11 Fuel

Attachment 2 to PLA-7947 Page 1 of 3 BACKGROUND In accordance with 10 CFR 50.46(a)(3)(ii), this annual report summarizes the nature of and estimated effect of any modeling changes or error cmrections in the ECCS model for the period January 21, 2021, through April 9, 2021. Additionally, further changes to the model are described herein which were identified from June 30, 2019, through January 21, 2021, while AURORA-B was being evaluated by the NRC for use at SSES, Units 1 and 2.

DISCUSSION ATRIUM 10 and ATRIUM 11 fuel are currently co-loaded in the SSES Unit 2 Reactor.

ATRIUM 10 is cmrently loaded in the SSES Unit 1 Reactor; ATRIUM 10 and ATRIUM 11 are expected to be co-loaded in the SSES Unit 1 Reactor following the refueling outage in spring 2022. The ECCS performance evaluation method applicable to ATRIUM 11 fuel for SSES Unit 2 is the Framatome AURORA-B LOCA (Loss of Coolant Accident) Methodology.

For the period of June 30, 2019, through April 9, 2021, there have been ten reportable 10 CFR 50.46 modeling changes or error corrections to the ECCS evaluation method.

The errors and changes have been captured in Framatome's Corrective Action Program.

These items do not meet the significance threshold for change (50°F) identified in 10 CFR 50.46(a)(3)(i) for which a 30-day report is required.

An error was identified in the Pellet-Cladding Mechanical Interaction routines in RODEX4. RODEX4 is a thermal-mechanical code used to calculate the1mal-mechanical properties of fuel rods. The impact of this change is summarized in Table 2.

(Reference 2)

An e1ror was identified in the way axial power shapes have been employed.

AUTOSR5LOCA is an automation code used to prepare inputs which are then used in the break spectrum analysis. A coding issue was identified during the process of generating axial power shapes associated with all heated core structures. This error is estimated to have zero impact on PCT. (Reference 2)

A change was identified in Susquehanna ATRIUM 11 lower tie plate bypass flow hole size which was revised subsequent to the ATRIUM 11 LOCA analysis. The bypass flow hole size is important to LOCA analysis since reverse flow through the hole affects reflood time. The impact due to the increased hole size is summarized in Table 2. (Reference 2)

An en-or was identified in the "pcmi_open_gap_aoo" model ofRODEX4. This error resulted in RODEX4 incorrectly assessing the gap size during a LOCA transient due to variables not being updated at the appropriate step time. RODEX4 is a thermal-

Attachment 2 to PLA-7947 Page 2 of 3 mechanical code used to calculate thermal-mechanical properties of fuel rods. This error is estimated to have zero impact on PCT. (Reference 3)

An error was identified in S-RELAP5 associated with the application of the Tien model. This error resulted in S-RELAP5 not applying the correct Tien model correction factor for the upper tie plate. S-RELAP5 is a thermal-hydraulic transient analysis program. This e1ror is estimated to have zero impact on PCT. (Reference 3)

An error was identified in AUTOBLT where the reference pressure location for safety relief valve operation was inconsistent for composite steam lines as compared to models with multiple steam lines. AUTOBLT is an automation code used to set up and execute LOCA calculations. This error is estimated to have zero impact on PCT.

(Reference 3)

An e1ror was identified in AUTOSR5BDK where the initialization flags for 2 of the 4 upper plenum to lower downcomer heat structures were identified to be inconsistent with the LOCA Topical Report sample problems. AUTOSR5BDK is a Calculation Processing Module designed to automate the preparation and stabilization of base input models for the S-RELAP5 thermal-hydraulic analysis program. This error is estimated to have zero impact on PCT. (Reference 3)

An error was identified in the upper tie plate (UTP) loss coefficient used in MICROBURN-B2 (UTP_LOSS_COEF) for the ATRIUM 11 fuel design. The approach used to remove the irreversible loss due to expansion was based on XCOBRA methods, not MICROBURN-B2. The impact of this change is summarized in Table 2.

(Reference 3)

An error was identified in the S-RELAP5 pump model. In the AURORA-B LOCA model the recirculation pump can go into reverse at high speeds for a significant amount of time which becomes dead ended after the recirculation line valve closes.

This results in umealistic energy deposition into an essentially steam filled space which can then reach umealistic temperatures resulting in case failures. This error is estimated to have zero impact on PCT. (Reference 4)

An error was identified that the RODEX4 kernel used in S-RELAP5 was failing because the code could reach a non-convergence iteration limit when it should not have.

The issue was resolved by setting the non-convergence iteration limit to zero at the start of every axial node calculation for each time step. This error is estimated to have zero impact on PCT. (Reference 4)

Attachment 2 to PLA-7947 Page 2 of 3 The total change listed in the last column of Table 2 does not meet the significance threshold for change (50°F) identified in 10 CFR 50.46(a)(3)(i) for which a 30-day report is required.

Attachment 2 to PLA-794 7 Page 3 of 3 IMPACT Table 2 Non-Zero PCT Changes Resulting from Modeling Changes/ Error Corrections in Calculated ECCS Performance Evaluation Model: Framatome AURORA-B LOCA Methodology Estimated Absolute APCT(°F) Value of Description of Change/En-or L\PCT (°F)

Pellet-Cladding Mechanical Interaction routines in RODEX4

-1 1 (Reference 2)

Lower tie plate bypass flow hole size increase

-3 3 (Reference 2)

Upper tie plate loss coefficient used in MICROBURN-B2

+5 5 (Reference 3)

Total Since Initial PCT

+1 9 (Reference 4)

CONCLUSION As documented in Table 2, the SSES Unit 2 Loss of Coolant Accident Analysis PCT remains in compliance with 10 CFR 50.46(b)(l), which requires that the PCT not exceed 2200°F.